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{{#Wiki_filter:1Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 1 of 23 Indian Point 2 Initiating Events Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
{{#Wiki_filter:}}
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 2 of 23 Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 3 of 23 Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 4 of 23 The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                                Page 5 of 23 Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 6 of 23 found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 7 of 23 Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 8 of 23 Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                                Page 9 of 23 Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 10 of 23 provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:          Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 11 of 23 undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 12 of 23 Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 13 of 23 Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 14 of 23 Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                      Page 15 of 23 Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 16 of 23 The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 17 of 23 Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 19 of 23 inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 20 of 23 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises,
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                      Page 21 of 23 commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 22 of 23 Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
 
1Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 23 of 23 Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Last modified : April 01, 2002
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 1 of 23 Indian Point 2 Initiating Events Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 2 of 23 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 3 of 23 a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Mitigating Systems Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 4 of 23 Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 5 of 23 Inspection Report# : 2001014(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                                Page 6 of 23 The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 7 of 23 Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 8 of 23 breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 9 of 23 FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR,
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 10 of 23 and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 11 of 23 Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 12 of 23 Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 13 of 23 EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 14 of 23 had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 15 of 23 Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 16 of 23 XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 17 of 23 Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Barrier Integrity Significance:          Feb 09, 2002 Identified By: NRC
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 19 of 23 Emergency Preparedness Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 20 of 23 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 21 of 23 Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 22 of 23 Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance:          Sep 30, 2000
 
2Q/2000 Inspection Findings - Indian Point 2                                                                                      Page 23 of 23 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:        Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Last modified : April 01, 2002
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 1 of 23 Indian Point 2 Initiating Events Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 2 of 23 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 3 of 23 a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Mitigating Systems Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 4 of 23 Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 5 of 23 the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                                Page 6 of 23 Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:          Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 7 of 23 Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 8 of 23 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 9 of 23 The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 10 of 23 Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 11 of 23 Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 12 of 23 a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 13 of 23 Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 14 of 23 analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 15 of 23 Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 16 of 23 The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 17 of 23 Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Barrier Integrity Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 19 of 23 Emergency Preparedness Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 20 of 23 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 21 of 23 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A,
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 22 of 23 Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A
 
3Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 23 of 23 selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Last modified : March 29, 2002
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 1 of 23 Indian Point 2 Initiating Events Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 2 of 23 Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 3 of 23 subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Mitigating Systems Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 4 of 23 three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 5 of 23 Inspection Report# : 2000011(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 6 of 23 analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 7 of 23 Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                                Page 8 of 23 The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 9 of 23 Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:          Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 10 of 23 Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 11 of 23 Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 12 of 23 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 13 of 23 Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 14 of 23 Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 15 of 23 analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 16 of 23 Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:        Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                              Page 17 of 23 Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Barrier Integrity Significance:          Jul 20, 2000 Identified By: NRC
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 19 of 23 Emergency Preparedness Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 20 of 23 Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 21 of 23 Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                          Page 22 of 23 The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000
 
4Q/2000 Inspection Findings - Indian Point 2                                                                                        Page 23 of 23 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Last modified : March 28, 2002
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 1 of 23 Indian Point 2 Initiating Events Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 2 of 23 Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 3 of 23 corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Mitigating Systems Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 4 of 23 Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:          Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 5 of 23 temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 6 of 23 Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 7 of 23 Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance: N/A Jan 13, 2001
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 8 of 23 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 9 of 23 specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 10 of 23 Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 11 of 23 operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Significance:        Feb 09, 2002
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 12 of 23 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 13 of 23 Significance:          Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 14 of 23 Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 15 of 23 Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 16 of 23 significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 17 of 23 Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Barrier Integrity Significance:          Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 19 of 23 Emergency Preparedness Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 20 of 23 determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 21 of 23 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A,
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 22 of 23 Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 23 of 23 Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Last modified : March 28, 2002
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 1 of 23 Indian Point 2 Initiating Events Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 2 of 23 transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 3 of 23 Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 4 of 23 of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 5 of 23 license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 6 of 23 Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:          Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 7 of 23 Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 8 of 23 Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 9 of 23 Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:        Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 10 of 23 specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 11 of 23 Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 12 of 23 The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:          Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 13 of 23 Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers,
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 14 of 23 but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 15 of 23 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 16 of 23 changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 17 of 23 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jul 20, 2000 Identified By: NRC
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 18 of 23 Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 19 of 23 Emergency Preparedness Significance:          Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 20 of 23 The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                      Page 21 of 23 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 22 of 23 to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000
 
2Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 23 of 23 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Last modified : March 27, 2002
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 1 of 22 Indian Point 2 Initiating Events Significance:        Feb 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 2 of 22 transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 3 of 22 Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:        Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 4 of 22 Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 5 of 22 completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 6 of 22 Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 7 of 22 reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 8 of 22 Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:          Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 9 of 22 temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:          Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 10 of 22 Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 11 of 22 of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 12 of 22 Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 13 of 22 XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 14 of 22 Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 15 of 22 Inspection Report# : 2001010(pdf)
Significance:          Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 16 of 22 Inspection Report# : 2000011(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 17 of 22 could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 18 of 22 Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 19 of 22 resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 20 of 22 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 21 of 22 Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 22 of 22 A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Last modified : March 26, 2002
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 1 of 20 Indian Point 2 Initiating Events Significance:        Dec 29, 2001 Identified By: Self Disclosing Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*,
"Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388)
(NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 2 of 20 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Jan 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 3 of 20 activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams.
The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 4 of 20 Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems.
Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution.
This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: TBD Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS The examiner determined that the crew high failure rate during facility administered annual NRC requalification exams had substantial safety significance. The crew failure is more than minor (credible effect on safety) because the rate is greater than 20% and the deficiencies identified during the exams reflected the potential inability of the crew to take appropriate safety related actions in response to actual abnormal or emergency conditions. The issue had substantial safety significance because of the multiple crew failures in that four of seven crews (57%) failed to meet Entergy requalification program requirements.
Inspection Report# : 2001013(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)
(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:          Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 5 of 20 Significance:          Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance.
Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of procedural guidance and maintenance of mitigating equipment for internal floods The inspector observed the flood door flaps located in the auxiliary feedwater pump room and the lower elevation of the primary auxiliary building could be hard to operate due to mechanical interference. The function of the door flaps is to swing open to direct flood water away from the auxiliary feedwater pumps and the residual heat removal pumps. This mitigation strategy is credited in IPEEE Section 5.0. The licensee documented this observation in CR 200108027. The inspector identified a difference between licensee commitments and the analysis in the IPEEE for a major flood within the turbine building. The NRC safety evaluation report (SER) concludes that design features and operating procedures provide assurance that the plant can be safely shutdown in the event of flooding outside containment from a non-seismic component or pipe. The issues are considered unresolved pending further NRC review to determine whether 1) operator actions within AOI 28.0.4 are adequate to mitigate a flood in the turbine building, and 2) the door flaps are functional to mitigating a postulated flood within the primary auxiliary building and auxiliary feed pump building. (UNR 05000247/2001-08-01)
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 6 of 20 resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293.
The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs).
There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 7 of 20 Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:          May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: URI Unresolved item AUXILIARY FEEDWATER SYSTEM DESIGN BASIS Although the inspector verified that operation of the TDAFW pump was in accordance with the UFSAR and other supporting documentation, additional NRC assessment was ongoing at the end of the inspection period. For example, although the MDAFW pump, as tested, provides adequate flow, based on the information provided the inspector was not able to determine that the AFW system could automatically provide sufficient cooling of post accident decay and sensible heat while delivering the minimum rated MDAFW pump flow indicated in the UFSAR. Further NRC review is required to determine the adequacy of the normal AFW system alignment with respect to its response to a feedline rupture. This issue is unresolved.
Inspection Report# : 2001004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                                Page 8 of 20 The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Gas turbine-2 became inoperable due to loss of air pressure During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor DRF to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Safeguards DC Power Failure Alarm The operators identified a failed status light on the train "A" blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the technical specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow tagging controls - CST inventory loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room operators if operations of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event could not result in a loss of safety
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 9 of 20 function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance.
This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
NCV 2001-003-02 Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 10 of 20 With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 11 of 20 analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                              Page 12 of 20 open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards,"
paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs)
- 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995.
The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 13 of 20 violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC.
Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures.
The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 14 of 20 Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:          Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 15 of 20 Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected.
The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                            Page 16 of 20 Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 17 of 20 a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)
(14).
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 18 of 20 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47 (b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                        Page 19 of 20 Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a),
Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
 
4Q/2001 Inspection Findings - Indian Point 2                                                                                          Page 20 of 20 Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Last modified : March 01, 2002
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 1 of 27 Indian Point 2 Initiating Events Significance:        Mar 30, 2002 Identified By: NRC Item Type: FIN Finding INAPPROPRIATE PROCEDURE FOR INOPERABLE STATION AUXILIARY TAP CHANGER The procedure in use was inappropriate in that it did not require that the 138 kilovolt off-site power system be declared inoperable during scheduled maintenance on the station auxiliary transformer (SAT) tap changer. On February 28, 2002, for approximately 51 minutes, control room operators had placed the SAT tap changer in manual and local control in accordance with system operating procedure (SOP) 27.1.7, "Operation of Main, Station and Unit Auxiliary Transformers," section 4.8. The scheduled maintenance was not intrusive into tap changer operation, however, the licensee had not fully evaluated if the intended function could be maintained with operator compensatory actions to restore the tap changer to automatic. The limiting condition for operation in technical specification 3.7.B.3 for a loss of the 138 kilovolt power system is 24 hours, which was not exceeded during this scheduled maintenance activity. The issue had a credible impact on safety. Inappropriate control of the SAT tap changer impacts the initiating event cornerstone in that a loss of off-site power is more likely following a reactor trip. This issue was determined to be of very low safety significance (Green) using phase one of the SDP because no reactor trip occurred during the inspection period and no mitigating systems were directly impacted by the maintenance on the SAT tap changer.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES FOR SW VALVE LOCKING DEVICES A personnel error was identified. One example of the cross cutting issue in human performance was the failure to properly maintain locking devices on five service water test stop valves. The failure to maintain locking devices on service water valves per the operating procedures was a violation of Technical Specification 6.8.1.a. This is a non-cited violation.
Inspection Report# : 2002002(pdf)
Significance:        Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR REPEAT FAILURE OF SWN-7 The manual operator on service water (SW) valve SWN-7 failed on March 9 during operations to swap essential SW headers. SWN-7 is the isolation valve for the service water supply to turbine building loads. The inoperable valve could have resulted in insufficient service water flow for the emergency diesel generators and other safety systems had there been a demand for those safety systems. The operator on SWN-7 failed 6 other times since 1995. Following the early failures, an engineering evaluation determined that the design margin for the gear box in the manual operator was marginally adequate. Engineering work request 12110-99 was issued to replace the gear set on SWN-7 and similar valves with high strength materials. The engineering request was canceled in July 1999 and no action was taken. This issue had very low safety significance since the specific failure on March 9 and corrective actions occurred within the limiting condition for operation for the service water system, and no operating or stand-by mitigating equipment supported by service water was called to perform its intended function. The failure to take adequate corrective action for repeat failures of service water valve SWN-7 was a violation of 10 CFR 50, Appendix B, Criterion XVI. This is being treated as a Non-cited violation.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation 10 CFR 50 APPENDIX B, CRITERION III, "DESIGN CONTROL" 10 CFR 50 Appendix B, Criterion III requires in part, that measures be established for the identification and control of design interfaces and for coordination among participating design organizations. The licensee did not ensure that the pressurizer level instrument drift evaluations were consistently bounded by the assumed instrument uncertainty within the safety analysis for a postulated Loss of Normal Feedwater event and a Loss of Offsite power event. The licensee documented this issue in condition report 2002000313.
Inspection Report# : 2002002(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 2 of 27 Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems. The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388) (NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Dec 29, 2001 Identified By: NRC Item Type: FIN Finding REACTOR TRIP AND PLANT RESPONSE On December 26, 2001, the reactor was automatically shutdown in response to a trip of the main turbine. The plant trip was caused by the failure of a non-safety related protection relay following a disturbance in the 345 KV electrical system that resulted in a partial load reject of the main generator output. The plant response was complicated by the de-energization of 6.9 KV buses 1 through 4, resulting in the shutdown of all four reactor coolant pumps, the de-energization of two of four 480 volt safeguard buses (safety buses 2A and 3A), and a loss of some of the operating condensate and circulating water pumps. The trip response was further complicated by equipment problems that resulted in the loss of the main condenser. For the fault that occurred in the 345 KV electrical system, the plant electrical response was as expected in accordance with the plant design. The licensee post trip evaluation demonstrated that turbine and reactor limits were not exceeded. The operators responded properly to the trip and the equipment performance problems. In accordance with NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because a reactor trip is a transient initiator and the plant transient with electrical complications could be a significant safety concern if the lost safety equipment was not readily recovered.
When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no impact on the plant safety barriers and the impact on mitigating safety equipment availability was minimal.
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment.
The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 3 of 27 Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance:        Jul 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination Event date was changed so that this item would show up during the ROP year 2002. The original date was July 20, 2000. The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects. Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws.
Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable. Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Inspection Report# : 2000011(pdf)
Significance:        May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding REDUCTION OF PLANT POWER BY CONTROL ROOM OPERATORS DUE TO CONDENSATE PUMP MOTOR FAILURES On April 20, 2002, and on May 8, 2002, the control room operators reduced plant power due to condensate pump motor failures. A lack of a predictive maintenance program and an improperly set oil level indication system were the causes for two separate condensate motor failures. The events are more than minor since both events increased the likelihood of an initiating event.
Operator response was necessary to ensure an automatic reactor trip did not occur due to a low steam generator level. The performance issues were of very low safety significance since there was no impact to normally available mitigating equipment.
Inspection Report# : 2002003(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 4 of 27 Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL IN TFC FOR NITROGEN BACKUP SYSTEM The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 5 of 27 An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams. The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance:          Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Problems with the auxiliary feedwater system during plant shutdown for mid-cycle maintenance outage During the plant shutdown for a mid-cycle maintenance outage on October 27, 2001, the operators experienced several problems with the auxiliary feedwater (AFW) system, which caused them to declare two motor driven pumps inoperable. Even though the auxiliary feedwater pumps were subsequently found to have been able to perform their intended safety function, the equipment operating deficiencies had a credible impact on the availability of the auxiliary feedwater system. The issue was evaluated in phase 1 of the SDP and was found to have very low safety significance.
Inspection Report# : 2001010(pdf)
Significance:          Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE. Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems. Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:          Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 6 of 27 condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented. Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution. This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:        Nov 05, 2001 Identified By: NRC Item Type: FIN Finding PROPOSED YELLOW FINDING DUE TO HIGH CREW FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with crew high failure rate (four of seven crews failed) during facility-administered annual licensed operator requalification examinations conducted last fall. The finding was previously characterized as having substantial safety significance (Yellow) in NRC Inspection Report 50-247/01-13. The inspectors noted that the licensee's evaluation identified a fundamental underlying weakness: The station has yet to overcome cultural weaknesses that include an unwillingness to confront poor performance, an over reliance on procedures to change behavior, and compartmentalization. More specifically, the licensee identified three root causes: 1) Operations training had not focused on the basic building blocks that ensure a healthy program; 2) The station had not maintained a core of career oriented, plant knowledgeable instructors and operators; and 3) Operations department involvement with Operations Training had often been ineffective. The inspectors concluded that the methodology and level of detail of the licensee's root cause evaluation were reasonable. The licensee implemented a number of corrective actions to address the identified causes. The corrective actions are described in the station's Training Improvement Plan. The more significant corrective actions included initiatives that aimed to 1) improve the quality of training and training materials; 2) increase the number of instructors who have Unit 2 plant experience; and 3) provide additional management support and oversight of training. The inspectors determined that the corrective actions are appropriately focused on the identified causes. These actions were appropriately prioritized, and either complete or scheduled for completion. Notably, the licensee took strong immediate corrective actions following the requalification examination failures to provide extensive retraining to each shift, and continue to provide this high intensity training. The inspectors independently assessed the extent of the underlying conditions that led to the Yellow finding and found that performance issues had also existed in other Operations Training programs, such as initial licensed operator and non-licensed operator training programs. These problems existed for at least three years, both prior to and following the steam generator tube failure event in 2001. Although licensee audits and assessments had identified most of the performance problems prior to the crew high failure rate, they did not identify long-term operator performance as a concern. The inspectors concluded that the licensee's extent of condition review appropriately bounded the underlying conditions that led to the Yellow finding as evidenced by the fact that the licensee had also identified the duration and extent of the problems, and the failure to recognize the long standing issues.
Inspection Report# : 2001013(pdf)
Inspection Report# : 2002004(pdf)
Inspection Report# : 2002009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: URI Unresolved item Reporting Safety System Functional Failures in PI Data Licensee event report 05000247/2000-006 documented that both source range instrument channel trip setpoints were outside the design basis due to the failure to account for postulated worst case ambient temperatures in the control room. Entergy did not classify this event as a safety system functional failure because the source range high flux trip is not credited in the UFSAR Chapter 14 accident analysis. The source range nuclear instruments are required to be operable per the technical specifications. NUREG-1022, Section 3.2.7, states that a failure of any component listed in the technical specification to perform a safety function, including shutdown of the reactor, is considered reportable under in 10 CFR 50.73(a)(2)(v). Further, if reported under this criteria, the failure would then meet the definition of a safety system functional failure. This item is considered unresolved pending further review by the NRC (UNR 05000247/01-09-01).
Inspection Report# : 2001009(pdf)
Significance:        Aug 18, 2001
 
1Q/2002 Inspection Findings - Indian Point 2                                                                                Page 7 of 27 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required). Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2.
This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance. Operator actions would be necessary to restore power to two of four safeguards buses. Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load.
Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 8 of 27 Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance:        Jul 07, 2001 Identified By: Self Disclosing Item Type: FIN Finding Failure of a fire water header.
During a test of the fire water system on June 5, 2001, the 12 inch fire water header failed, which resulted in a leak of 231,000 gallons of city water into the Utility Tunnel. The automatic and manual fire suppression system was inoperable for approximately 1 hour and 15 minutes, which impacted 14 fire zones that contained alternate safe shutdown equipment. The licensee restored the main fire header back to a fully functional status on June 10, 2001. The fire header failed because of inadequate alignment and torque setting of the Victaulic couplings when the header was modified in November 2000. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 9 of 27 retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293. The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs). There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance:        May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 10 of 27 Significance:          Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:          Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited.
This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators.
 
1Q/2002 Inspection Findings - Indian Point 2                                                                                  Page 11 of 27 Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply The operators identified a failed status light on the train A blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the Technical Specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Tagging Controls - CST Inventory Loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room if operation of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event did not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance. This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding GT-3 found inoperable while GT-2 was out of service for corrective maintenance and EDG 22 were out of service for planned maintenance During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-
: 11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal. Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern. For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03
 
1Q/2002 Inspection Findings - Indian Point 2                                                                          Page 12 of 27 Inspection Report# : 2001003(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means.
The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:        Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 13 of 27 Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs) - 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995. The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows.
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 14 of 27 However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve.
The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 15 of 27 Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards," paragraph (f), "Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:          Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 16 of 27 Significance:        Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function.
The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC. Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90%
(89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown. The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours.
However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures. The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was
 
1Q/2002 Inspection Findings - Indian Point 2                                                                          Page 17 of 27 in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted.
Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures.
The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:        Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 18 of 27 Significance:        Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 19 of 27 measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected. The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 20 of 27 was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance:        May 11, 2002 Identified By: Licensee Item Type: FIN Finding UNTIMELY OPERATOR EVALUATION FOR CONTAINMENT ISOLATION VALVE 869B On April 11, 2002, operators did not complete a timely operability evaluation for containment isolation valve 869B after the disconnect switch operating handle on motor control center (MCC)26BB broke while applying an equipment tagout. At the time, the operators neither verified that the disconnect would operate nor completed an adequate evaluation regarding the ability to close valve 869B to perform its containment isolation function. An operability evaluation was completed about six hours later by a different operating crew and the operators then entered a four-hour limiting condition for operation and isolated the containment penetration per the technical specifications 3.6.A.3.a.2.b. The untimely operability evaluation increased the unavailability time for the containment spray system. The inoperable containment isolation valve issue was more than minor because it impacts the containment barrier. This issue had very low safety significance since the containment isolation valve was repaired and restored to an operable status prior to exceeding technical specification 3.6.A.3.a.2.d. This issue was an example of untimely operator implementation of technical specification requirements in response to degraded safety equipment.
Inspection Report# : 2002003(pdf)
Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-
qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000.
While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance:        Jun 25, 2001
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 21 of 27 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan. The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance: N/A May 11, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF 10 CFR 50.54(q) FOR ACCOUNTABILITY On March 6, 2002, the licensee implemented changes to the accountability process that decreased the effectiveness of the Emergency Plan (E-Plan). The finding was considered more than minor because, if left uncorrected, it would become a more significant safety concern. Changing commitments in the E-Plan without prior approval impacts the NRC's ability to perform its regulatory function and potentially creates an ineffective response to a radiological emergency. The consequences of this change were minimal because it did not preclude the function of accountability from being performed, albeit delayed. The licensee has implemented the corrective actions and has since met the timeliness goal. This change which decreased the effectiveness of the Plan is being treateed as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued May 1, 2000.
Inspection Report# : 2002003(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)(14).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 22 of 27 Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse.
This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:          Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47(b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:          Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 23 of 27 Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector. Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a), Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 24 of 27 documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD)
Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance: N/A May 11, 2002 Identified By: Licensee Item Type: NCV NonCited Violation VIOLATION OF TECHNICAL SPECIFICATION 6.8.1.a - IMPROPER PROCEDURE USAGE On April 20, 2002, during a trip of one of the three condensate pumps, control room operators took incorrect action based on an abnormal operating instruction (AOI 21.1.1 step 5.6.4, by using a suction pressure number from this step that did not apply which resulted in their taking operator actions resulting in an unnecessary power transient. A May 8, 2002 condensate pump trip exemplified that this transient (a rapid down power) was not necessary to restore feedwater pump suction. The issue was more than minor since operator improper procedure usage is considered a precursor to a more significant event. Operator knowledge and skill performance issues have been captured in a number of individual NRC findings in past reports. Examples include operator re-qualification simulator test failures in September 2001 (reference NRC report 50-247/2001-013), and an overpower condition in August 2001 (reference NRC report 50-247/2001-09). The operator performance issues associated with the condensate pump trip were documented in the corrective action system as CRs 2000204180 and 200204183. Improper AOI 21.1.1 procedure usage was a violation of Technical Specification 6.8.1.a. This is being treated as a non-cited violation.
Inspection Report# : 2002003(pdf)
Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves.
This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 25 of 27 Significance:          Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:          May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Conclusions for Maintenance and Surveillance Maintenance activities were satisfactorily completed. The conduct of surveillance tests during the period was acceptable.
Maintenance and test activities were not consistently performed in accordance with expectations and administrative controls. The initial evaluations in preparation for a turbine load test did not completely consider shutdown risk.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Container Liner Degradation The containment liner became corroded due to prolonged contact with borated water in areas where moisture barriers were degraded. Con Edison actions continued to investigate and repair liner degradation, and to assure that margins to design limits were maintained.
 
1Q/2002 Inspection Findings - Indian Point 2                                                                              Page 26 of 27 Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Steam Generator Examinations Steam generator eddy current testing and analysis was conducted. The eddy current test results revealed defects which resulted in a Classification of C-3 per Technical Specifrication 4.13. More detailed review of steam generator inspecction results is under the purview of the NRC Office of Nuclear Reactor Regulation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Nuclear Facilities Safety Committee Plant management presentations to the Nuclear Facilities Safety Committee were incomplete. However, the committee members appeared well prepared and provided good discussions on the February 15 steam generator tube leak event.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak. Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Review of Response to Loss of Power and Air Supply to Steam Generator Nozzle Dams The operators promptly responded to the loss of power to the steam generator nozzle dams. The nozzle dam normal air supply was lost; however, no loss of reactor coolant system inventory occurred, and no monitoring existed for the nozzle dams for approximately one hour. Con Edison failed to control and integrate several temporary facility changes for the nozzle dam support systems.
Inadequate coordination between operators and workers resulted in a near miss for a significant injury Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES ON CORE DIFFERENTIAL TEMPERATURE The operators failed to control RCS differential temperature within limits during RHR system operation. The failure to follow SOP 4.2.1 was a non-cited violation of NRC requirements. Licensee actions continued at the end of the inspection period to evaluate the impact on the baffle-former and baffle-barrel bolts in the reactor vessel internals, and to resolve this matter prior to plant restart.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET APPENDIX R FOR OIL COLLECTION SYSTEM The failure to collect leakage from the vent pipe and the lower oil reservoir drain connections on three RCP motors is considered a violation of 10 CFR 50 Appendix R, Section III. This Severity Level IV violation is being treated as a Non-Cited Violation. A long-standing deficiency in the oil collection system went uncorrected.
 
1Q/2002 Inspection Findings - Indian Point 2                                                                            Page 27 of 27 Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: AV Apparent Violation FAILURE TO MEET IVSWS LICENSING BASIS Con Edison did not recognize a long-standing difference between the design and licensing basis for the isolation valve seal water system. Despite several past events and a design basis verification program which highlighted IVSWS performance issues, Con Edison failed to correct a basic design deficiency and assure that the licensing basis was met. Operability evaluations were less than adequate and corrective actions were narrow and untimely. The failure to assure regulatory requirements were correctly translated into specifications, drawings and procedures was an apparent violation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE INSTRUCTIONS FOR FIRE DAMPERS lack of maintenance installation instructions contributed to the failure of cable spreading room fire dampers to fully close. The faulty dampers caused the suppression system to be degraded for approximately 3 months. The failure to maintain provisions of the NRC-approved fire protection plan as described in the UFSAR and approved NRC Safety Evaluation Report is a Non-Cited Violation.
Inspection Report# : 2000003(pdf)
Last modified : July 22, 2002
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 1 of 35 Indian Point 2 Initiating Events Significance:      May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding REDUCTION OF PLANT POWER BY CONTROL ROOM OPERATORS DUE TO CONDENSATE PUMP MOTOR FAILURES On April 20, 2002, and on May 8, 2002, the control room operators reduced plant power due to condensate pump motor failures. A lack of a predictive maintenance program and an improperly set oil level indication system were the causes for two separate condensate motor failures. The events are more than minor since both events increased the likelihood of an initiating event. Operator response was necessary to ensure an automatic reactor trip did not occur due to a low steam generator level. The performance issues were of very low safety significance since there was no impact to normally available mitigating equipment.
Inspection Report# : 2002003(pdf)
Significance:      Mar 30, 2002 Identified By: NRC Item Type: FIN Finding INAPPROPRIATE PROCEDURE FOR INOPERABLE STATION AUXILIARY TAP CHANGER The procedure in use was inappropriate in that it did not require that the 138 kilovolt off-site power system be declared inoperable during scheduled maintenance on the station auxiliary transformer (SAT) tap changer. On February 28, 2002, for approximately 51 minutes, control room operators had placed the SAT tap changer in manual and local control in accordance with system operating procedure (SOP) 27.1.7, "Operation of Main, Station and Unit Auxiliary Transformers," section 4.8. The scheduled maintenance was not intrusive into tap changer operation, however, the licensee had not fully evaluated if the intended function could be maintained with operator compensatory actions to restore the tap changer to automatic. The limiting condition for operation in technical specification 3.7.B.3 for a loss of the 138 kilovolt power system is 24 hours, which was not exceeded during this scheduled maintenance activity. The issue had a credible impact on safety. Inappropriate control of the SAT tap changer impacts the initiating event cornerstone in that a loss of off-site power is more likely following a reactor trip. This issue was determined to be of very low safety significance (Green) using phase one of the SDP because no reactor trip occurred during the inspection period and no mitigating systems were directly impacted by the maintenance on the SAT tap changer.
Inspection Report# : 2002002(pdf)
Significance:      Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 2 of 35 inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems.
The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388) (NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:        Jul 01, 2001 Identified By: NRC Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination Event date was changed so that this item would show up during the ROP year 2002. The original date was July 20, 2000. The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects.
Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 3 of 35 significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable.
Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Inspection Report# : 2000011(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:        May 26, 2000 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 4 of 35 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:      Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding MULTIPLE GROUNDS ON THE PROTECTIVE CIRCUIT FOR UNIT 1 SUBSTATION 102NS3 RESULTED IN A LOSS OF THE 13.8 KILOVOLT LIGHTING AND POWER BUS SECTION 3 On May 17, 2002, multiple grounds on the protective circuit for Unit 1 substation 102NS3 resulted in a loss of the 13.8 kilovolt (kv) lighting and power bus section 3. The consequence of this event was a loss of alternate safe shutdown power to all major alternate safe shutdown pumps and selected instrumentation. At the time, the Unit 2 normal and emergency electrical power supplies were available to supply power to the above stated mitigation equipment and instrumentation. The licensee repaired and restored the 13.8 kv bus section 3 within 30 hours of the fault. The performance issue is inadequate retirement of protective circuits for 440 volt substations (132PC3 and 142PC3) that could impact availability of alternate safe shutdown power supplies. This issue is more than minor since unavailability of alternate safe shutdown equipment for 30 hours is viewed as a precursor to a significant event and the alternate safe shutdown power supplies are a risk-significant maintenance rule system which was unavailable for greater than 24 hours.
Inspection Report# : 2002004(pdf)
Significance:      Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR A TEMPORARY FACILITY CHANGE INVOLVING THE AUXILIARY FEEDWATER SYSTEM BACKUP NITROGEN SUPPLY SYSTEM.
The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 5 of 35 to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems. Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:        Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:        Nov 05, 2001 Identified By: NRC Item Type: FIN Finding CREW HIGH FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with crew high failure rate (four of seven crews failed) during facility-administered annual licensed operator requalification examinations conducted last fall. The finding was previously characterized as having substantial safety significance file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 6 of 35 (Yellow) in NRC Inspection Report 50-247/01-13. The inspectors noted that the licensee's evaluation identified a fundamental underlying weakness: The station has yet to overcome cultural weaknesses that include an unwillingness to confront poor performance, an over reliance on procedures to change behavior, and compartmentalization. More specifically, the licensee identified three root causes: 1) Operations training had not focused on the basic building blocks that ensure a healthy program; 2) The station had not maintained a core of career oriented, plant knowledgeable instructors and operators; and 3) Operations department involvement with Operations Training had often been ineffective. The inspectors concluded that the methodology and level of detail of the licensee's root cause evaluation were reasonable. The licensee implemented a number of corrective actions to address the identified causes. The corrective actions are described in the station's Training Improvement Plan. The more significant corrective actions included initiatives that aimed to 1) improve the quality of training and training materials; 2) increase the number of instructors who have Unit 2 plant experience; and 3) provide additional management support and oversight of training.
The inspectors determined that the corrective actions are appropriately focused on the identified causes. These actions were appropriately prioritized, and either complete or scheduled for completion. Notably, the licensee took strong immediate corrective actions following the requalification examination failures to provide extensive retraining to each shift, and continue to provide this high intensity training. The inspectors independently assessed the extent of the underlying conditions that led to the Yellow finding and found that performance issues had also existed in other Operations Training programs, such as initial licensed operator and non-licensed operator training programs. These problems existed for at least three years, both prior to and following the steam generator tube failure event in 2001.
Although licensee audits and assessments had identified most of the performance problems prior to the crew high failure rate, they did not identify long-term operator performance as a concern. The inspectors concluded that the licensee's extent of condition review appropriately bounded the underlying conditions that led to the Yellow finding as evidenced by the fact that the licensee had also identified the duration and extent of the problems, and the failure to recognize the long standing issues. (Updated) FIN 05000247/01-013-01: Proposed finding due to crew high failure rate during the 2001 annual requalification simulator examinations. This finding was documented in an October 2001 inspection and initially characterized as a potential Yellow finding, the final safety significance to be determined (TBD). This finding was subsequently evaluated under the significance determination process (SDP) and characterized as (reference NRC to Entergy letters dated December 5, 2001, and February 28, 2002). The 95002 Supplemental Inspection (reference Inspection Report No. 50-247/02-09, dated May 31, 2002), assessed the licensee's evaluation of the crew high failure rates and the corrective actions taken to address this performance issue. As stated in the cover letter to Inspection Report No. 50-247/02-09, this finding remains open until after the completion of Entergy's licensed operator requalification examinations, scheduled for September-October 2002, and further review by the NRC. This item remains open.
Inspection Report# : 2001013(pdf)
Inspection Report# : 2002004(pdf)
Inspection Report# : 2002009(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit.
This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 7 of 35 Item Type: NCV NonCited Violation Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance. Operator actions would be necessary to restore power to two of four safeguards buses.
Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance:      Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance:      Jul 07, 2001 Identified By: Self Disclosing Item Type: FIN Finding Failure of a fire water header.
During a test of the fire water system on June 5, 2001, the 12 inch fire water header failed, which resulted in a leak of 231,000 gallons of city water into the Utility Tunnel. The automatic and manual fire suppression system was inoperable for approximately 1 hour and 15 minutes, which impacted 14 fire zones that contained alternate safe shutdown equipment. The licensee restored the main fire header back to a fully functional status on June 10, 2001. The fire header failed because of inadequate alignment and torque setting of the Victaulic couplings when the header was modified in November 2000. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001006(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 8 of 35 concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293. The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs). There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                    Page 9 of 35 testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                    Page 10 of 35 Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply The operators identified a failed status light on the train A blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the Technical Specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding GT-3 found inoperable while GT-2 was out of service for corrective maintenance and EDG 22 were out of service for planned maintenance During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 11 of 35 preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal.
Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern.
For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Tagging Controls - CST Inventory Loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room if operation of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event did not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance. This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
Inspection Report# : 2001003(pdf)
Significance:      Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 12 of 35 having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:      Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:      Feb 09, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                    Page 13 of 35 Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 14 of 35 to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 15 of 35 low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards," paragraph (f),
"Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs) - 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995. The team did an independent calculation of the change in core file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 16 of 35 damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                    Page 17 of 35 Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC. Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 18 of 35 below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown.
The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures. The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 19 of 35 Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:        Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 20 of 35 fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:      May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 21 of 35 that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:      May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected. The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:      Jun 29, 2002 Identified By: NRC Item Type: FIN Finding DURING SURVEILLANCE TESTING OF THE SAFETY INJECTION DISCHARGE MOTOR-OPERATED VALVE 851B, THE VALVE FAILED TO STROKE CLOSED On May 27, 2002, during surveillance testing of the safety injection discharge motor-operated valve (851B), the valve failed to stroke closed. The initial operability evaluation did not consider the non-automatic containment isolation function for this valve. This event was documented in condition report No. 200205433. The performance issue associated with this finding is a weakness in operator knowledge of multi-function safety system components. This is the second recent example where operators did not consider this function for a safety-related valve. The first example was documented in NRC report 50-247/2002-003, section 1R15. The untimely and incomplete operability assessment for safety injection discharge valve 851B has very low safety significance since the containment isolation valve was restored to an operable status prior to exceeding Technical Specification 3.6.A.3.a.2.d limiting condition for operation.
Inspection Report# : 2002004(pdf)
Significance:      May 11, 2002 Identified By: Licensee file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 22 of 35 Item Type: FIN Finding UNTIMELY OPERATOR EVALUATION FOR CONTAINMENT ISOLATION VALVE 869B On April 11, 2002, operators did not complete a timely operability evaluation for containment isolation valve 869B after the disconnect switch operating handle on motor control center (MCC)26BB broke while applying an equipment tagout. At the time, the operators neither verified that the disconnect would operate nor completed an adequate evaluation regarding the ability to close valve 869B to perform its containment isolation function. An operability evaluation was completed about six hours later by a different operating crew and the operators then entered a four-hour limiting condition for operation and isolated the containment penetration per the technical specifications 3.6.A.3.a.2.b.
The untimely operability evaluation increased the unavailability time for the containment spray system. The inoperable containment isolation valve issue was more than minor because it impacts the containment barrier. This issue had very low safety significance since the containment isolation valve was repaired and restored to an operable status prior to exceeding technical specification 3.6.A.3.a.2.d. This issue was an example of untimely operator implementation of technical specification requirements in response to degraded safety equipment.
Inspection Report# : 2002003(pdf)
Significance:      Feb 09, 2002 Identified By: Licensee Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:      Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 23 of 35 Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance:        Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity.
The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance: N/A May 11, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF 10 CFR 50.54(q) FOR ACCOUNTABILITY On March 6, 2002, the licensee implemented changes to the accountability process that decreased the effectiveness of the Emergency Plan (E-Plan). The finding was considered more than minor because, if left uncorrected, it would become a more significant safety concern. Changing commitments in the E-Plan without prior approval impacts the NRC's ability to perform its regulatory function and potentially creates an ineffective response to a radiological emergency. The consequences of this change were minimal because it did not preclude the function of accountability from being performed, albeit delayed. The licensee has implemented the corrective actions and has since met the timeliness goal. This change which decreased the effectiveness of the Plan is being treateed as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued May 1, 2000.
Inspection Report# : 2002003(pdf)
Significance:        Jun 25, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 24 of 35 Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan.
The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)(14).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 25 of 35 Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 26 of 35 Significance:      Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47(b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:      Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity.
This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:      Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:      Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 27 of 35 Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector.
Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a), Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance: N/A May 11, 2002 Identified By: Licensee Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 28 of 35 VIOLATION OF TECHNICAL SPECIFICATION 6.8.1.a - IMPROPER PROCEDURE USAGE On April 20, 2002, during a trip of one of the three condensate pumps, control room operators took incorrect action based on an abnormal operating instruction (AOI 21.1.1 step 5.6.4, by using a suction pressure number from this step that did not apply which resulted in their taking operator actions resulting in an unnecessary power transient. A May 8, 2002 condensate pump trip exemplified that this transient (a rapid down power) was not necessary to restore feedwater pump suction. The issue was more than minor since operator improper procedure usage is considered a precursor to a more significant event. Operator knowledge and skill performance issues have been captured in a number of individual NRC findings in past reports. Examples include operator re-qualification simulator test failures in September 2001 (reference NRC report 50-247/2001-013), and an overpower condition in August 2001 (reference NRC report 50-247/2001-09). The operator performance issues associated with the condensate pump trip were documented in the corrective action system as CRs 2000204180 and 200204183. Improper AOI 21.1.1 procedure usage was a violation of Technical Specification 6.8.1.a. This is being treated as a non-cited violation.
Inspection Report# : 2002003(pdf)
Significance: N/A Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES FOR SW VALVE LOCKING DEVICES A personnel error (human performance cross cutting issue) resulted in the failure to properly maintain locking devices on five service water test stop valves. The failure to maintain locking devices on service water valves per the operating procedures was a violation of Technical Specification 6.8.1.a. This is a non-cited violation.
Inspection Report# : 2002002(pdf)
Significance:        Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR REPEAT FAILURE OF SWN-7 The manual operator on service water (SW) valve SWN-7 failed on March 9 during operations to swap essential SW headers. SWN-7 is the isolation valve for the service water supply to turbine building loads. The inoperable valve could have resulted in insufficient service water flow for the emergency diesel generators and other safety systems had there been a demand for those safety systems. The operator on SWN-7 failed 6 other times since 1995. Following the early failures, an engineering evaluation determined that the design margin for the gear box in the manual operator was marginally adequate. Engineering work request 12110-99 was issued to replace the gear set on SWN-7 and similar valves with high strength materials. The engineering request was canceled in July 1999 and no action was taken. This issue had very low safety significance since the specific failure on March 9 and corrective actions occurred within the limiting condition for operation for the service water system, and no operating or stand-by mitigating equipment supported by service water was called to perform its intended function. The failure to take adequate corrective action for repeat failures of service water valve SWN-7 was a violation of 10 CFR 50, Appendix B, Criterion XVI. This is being treated as a Non-cited violation.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation 10 CFR 50 APPENDIX B, CRITERION III, "DESIGN CONTROL" 10 CFR 50 Appendix B, Criterion III requires in part, that measures be established for the identification and control of design interfaces and for coordination among participating design organizations. The licensee did not ensure that the pressurizer level instrument drift evaluations were consistently bounded by the assumed instrument uncertainty within the safety analysis for a postulated Loss of Normal Feedwater event and a Loss of Offsite power event. The licensee file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 29 of 35 documented this issue in condition report 2002000313.
Inspection Report# : 2002002(pdf)
Significance: N/A Feb 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams. The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 30 of 35 Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented.
Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution. This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                Page 31 of 35 addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required).
Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 32 of 35 While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:      Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance:      Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 33 of 35 Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Container Liner Degradation The containment liner became corroded due to prolonged contact with borated water in areas where moisture barriers were degraded. Con Edison actions continued to investigate and repair liner degradation, and to assure that margins to design limits were maintained.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 34 of 35 Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Conclusions for Maintenance and Surveillance Maintenance activities were satisfactorily completed. The conduct of surveillance tests during the period was acceptable. Maintenance and test activities were not consistently performed in accordance with expectations and administrative controls. The initial evaluations in preparation for a turbine load test did not completely consider shutdown risk.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Nuclear Facilities Safety Committee Plant management presentations to the Nuclear Facilities Safety Committee were incomplete. However, the committee members appeared well prepared and provided good discussions on the February 15 steam generator tube leak event.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Review of Response to Loss of Power and Air Supply to Steam Generator Nozzle Dams The operators promptly responded to the loss of power to the steam generator nozzle dams. The nozzle dam normal air supply was lost; however, no loss of reactor coolant system inventory occurred, and no monitoring existed for the nozzle dams for approximately one hour. Con Edison failed to control and integrate several temporary facility changes for the nozzle dam support systems. Inadequate coordination between operators and workers resulted in a near miss for a significant injury Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Steam Generator Examinations file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Indian Point 2                                                                  Page 35 of 35 Steam generator eddy current testing and analysis was conducted. The eddy current test results revealed defects which resulted in a Classification of C-3 per Technical Specifrication 4.13. More detailed review of steam generator inspecction results is under the purview of the NRC Office of Nuclear Reactor Regulation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES ON CORE DIFFERENTIAL TEMPERATURE The operators failed to control RCS differential temperature within limits during RHR system operation. The failure to follow SOP 4.2.1 was a non-cited violation of NRC requirements. Licensee actions continued at the end of the inspection period to evaluate the impact on the baffle-former and baffle-barrel bolts in the reactor vessel internals, and to resolve this matter prior to plant restart.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET APPENDIX R FOR OIL COLLECTION SYSTEM The failure to collect leakage from the vent pipe and the lower oil reservoir drain connections on three RCP motors is considered a violation of 10 CFR 50 Appendix R, Section III. This Severity Level IV violation is being treated as a Non-Cited Violation. A long-standing deficiency in the oil collection system went uncorrected.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: AV Apparent Violation FAILURE TO MEET IVSWS LICENSING BASIS Con Edison did not recognize a long-standing difference between the design and licensing basis for the isolation valve seal water system. Despite several past events and a design basis verification program which highlighted IVSWS performance issues, Con Edison failed to correct a basic design deficiency and assure that the licensing basis was met.
Operability evaluations were less than adequate and corrective actions were narrow and untimely. The failure to assure regulatory requirements were correctly translated into specifications, drawings and procedures was an apparent violation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE INSTRUCTIONS FOR FIRE DAMPERS lack of maintenance installation instructions contributed to the failure of cable spreading room fire dampers to fully close. The faulty dampers caused the suppression system to be degraded for approximately 3 months. The failure to maintain provisions of the NRC-approved fire protection plan as described in the UFSAR and approved NRC Safety Evaluation Report is a Non-Cited Violation.
Inspection Report# : 2000003(pdf)
Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/03/2003
 
3Q/2002 Inspection Findings - Indian Point 2                                                                    Page 1 of 36 Indian Point 2 Initiating Events Significance:        Aug 10, 2002 Identified By: Self Disclosing Item Type: FIN Finding CONTRACTOR WORKED OUTSIDE HIS ESTABLISHED JOB SCOPE FOR LANDSCAPING ACTIVITIES On July 19, 2002, a contractor worked outside his established job scope for landscaping activities. The consequences of this human performance error were the accidental electrocution of the individual and an offsite power electrical transient (loss of the 138 kilovolt station auxiliary transformer for approximately seven hours). This partial loss of offsite power event was more than minor, in that it impacted the reactor safety cornerstone with respect to the initiating event objective of limiting the likelihood of an event that upsets plant stability and challenges the critical safety function of the on-site emergency diesel generators. Notwithstanding the loss of life (which the Department of Labor, Occupational Safety and Health Administration is reviewing), this electrical transient event was of very low safety significance because it did not contribute to the likelihood of: loss of coolant accidents, a reactor trip and the unavailability of accident mitigation equipment or functions being unavailable; or of a fire or internal/external flood.
No violations of NRC requirements were identified.
Inspection Report# : 2002005(pdf)
Significance:        May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding REDUCTION OF PLANT POWER BY CONTROL ROOM OPERATORS DUE TO CONDENSATE PUMP MOTOR FAILURES On April 20, 2002, and on May 8, 2002, the control room operators reduced plant power due to condensate pump motor failures. A lack of a predictive maintenance program and an improperly set oil level indication system were the causes for two separate condensate motor failures. The events are more than minor since both events increased the likelihood of an initiating event. Operator response was necessary to ensure an automatic reactor trip did not occur due to a low steam generator level. The performance issues were of very low safety significance since there was no impact to normally available mitigating equipment.
Inspection Report# : 2002003(pdf)
Significance:        Mar 30, 2002 Identified By: NRC Item Type: FIN Finding INAPPROPRIATE PROCEDURE FOR INOPERABLE STATION AUXILIARY TAP CHANGER The procedure in use was inappropriate in that it did not require that the 138 kilovolt off-site power system be declared inoperable during scheduled maintenance on the station auxiliary transformer (SAT) tap changer. On February 28, 2002, for approximately 51 minutes, control room operators had placed the SAT tap changer in manual and local control in accordance with system operating procedure (SOP) 27.1.7, "Operation of Main, Station and Unit Auxiliary Transformers," section 4.8. The scheduled maintenance was not intrusive into tap changer operation, however, the licensee had not fully evaluated if the intended function could be maintained with operator compensatory actions to restore the tap changer to automatic. The limiting condition for operation in technical specification 3.7.B.3 for a loss of the 138 kilovolt power system is 24 hours, which was not exceeded during this scheduled maintenance activity. The issue had a credible impact on safety. Inappropriate control of the SAT tap changer impacts the initiating event cornerstone in that a loss of off-site power is more likely following a reactor trip. This issue was determined to be of
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 2 of 36 very low safety significance (Green) using phase one of the SDP because no reactor trip occurred during the inspection period and no mitigating systems were directly impacted by the maintenance on the SAT tap changer.
Inspection Report# : 2002002(pdf)
Significance:      Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation OPERATOR FAILURE TO PLACE MODE SWITCH TO AUTO RESULTING IN DILUTION OF THE RCS BY AN ADDITIONAL SIX GALLONS While making a routine RCS dilution on December 17, 2001, an operator error resulted in an inadvertent dilution of 6 additional gallons of primary water (a total of 42 gallons was added versus the 36 gallons planned). The error occurred because the operator failed to place the Mode switch to AUTO per Step 4.3.16(4) of SOP 3.2 when securing the CVCS from the Dilution mode. The failure to follow procedures was contrary to Technical Specification 6.8.1.a. The inadvertent RCS dilution was classified as a reactivity management event. In accordance with the NRC Manual Chapters 0609, "Significance Determination Process," and 0610*, "Power Reactor Inspection Reports," this issue was determined to be more that minor because an inadvertent dilution of the RCS, if left uncorrected, could become a more significant safety concern. When evaluated in accordance with the SDP Phase 1, the issue was considered to be of very low safety significance since there was no actual challenge to reactor safety or the status of mitigating safety systems.
The licensee identified this procedure violation (reference condition report 200112470). This failure to adhere to a procedure is being treated as a non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388) (NCV 50-247/01-11-01).
Inspection Report# : 2001011(pdf)
Significance:      Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Several Examples of Failure to Follow Calorimetric Procedure The operators' failure to follow calorimetric and operating procedures resulted in an overpower condition on August 17, 2001, and was a violation of Technical Specification 6.8.1. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of Technical Specification 6.8.1.a was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:      Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Poor Reactivity Management Caused Violation of Power Limit The operators' failure to adequately monitor plant conditions resulted in an overpower condition on August 17, 2001, and a violation of the License Condition 2.C.(1) thermal power limit. The overpower condition impacted the reactor safety cornerstone since it could have caused a reactor trip if not corrected by the operators. This event had very low safety significance, since the overpower condition was minor, existed for a small amount of time, and resulted in no loss of function or availability of mitigation equipment. The violation of License Condition 2.C.(1) was treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001009(pdf)
Significance:      Jul 01, 2001 Identified By: NRC
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 3 of 36 Item Type: VIO Violation Deficiencies in the overall direction and execution of the 1997 SG inservice examination Event date was changed so that this item would show up during the ROP year 2002. The original date was July 20, 2000. The overall direction and execution of the 1997 SG inservice examinations were deficient in several respects.
Despite opportunities, Con Edison did not identify and correct a significant condition adverse to quality involving the presence of primary water stress corrosion cracking (PWSCC) flaws in row 2 steam generator (SG) tubes in the small radius, low-row U-bend apex area. Con Edison did not adequately account for conditions which adversely affected the detectability of, and increased the susceptibility to, tube flaws. Specifically during the 1997 SG Eddy Current Test (ECT) and secondary side visual examination. As a result, tubes with PWSCC flaws in their small radius U-bends were left in service following the 1997 inspection, until the failure of these tubes occured on February 15, 2000, while the reactor was at 100-percent power. This preliminary finding was characterized as Red, an issue of high safety significance, in inspection report 05000247/2000-010, dated August 31, 2000. Final assessment of the inspection finding using the SDP was characterized as Red and provided to the licensee in a {{letter dated|date=November 20, 2000|text=letter dated November 20, 2000}}, subsequent to a regulatory conference that was held on September 26, 2000. The NRC determined that the licensee's failure to identify and adjust or modify the inspection methods and analysis to account for significant conditions that affected the quality of the 1997 steam generator inspection was a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions. In a {{letter dated|date=January 19, 2001|text=letter dated January 19, 2001}}, the licensee denied that the violation occurred and contended that the 1997 steam generator tube inservice examination was conducted in accordance with industry guidelines and requirements applicable at the time. The licensee also provided several affadavits prepared by individuals with experience in steam generator inspection and eddy current testing, attesting licensee performance to be acceptable.
Additional NRC review of the licensee's response and bases for denial of the violation did not alter the NRC's conclusion that the violation existed. NRC follow-up to this issue will focus on the licensee's corrective action program effectiveness.
Inspection Report# : 2000010(pdf)
Inspection Report# : 2000011(pdf)
Inspection Report# : 2002010(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Higher failure rate on the year 2000 requalification examinations The facility has experienced a high failure rate on the Year 2000 requalification examinations. This is attributable in part to an upgrade in examination difficulty. The significance of this issue is low; however, a high failure rate may indicate poor training and inadequate competence level. This did not appear to be the case because the facility had increased the difficulty level of the written examinations for their Year 2000 exams and exams administered in 1998 were adequate.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation NRC identified that the licensee did not sample all Senior Reactor Operators on emergency plan implementation The facility did not design their annual operating test such that all Senior Reactor Operator licensees were "at risk" of being evaluated on implementation of the emergency plan. The safety significance of this finding is low because emergency plan knowledge was tested on the written examination and sampled in the Year 2000 operational examinations after this inspection. This is a non-cited violation of 10CFR55.59(a)(2).
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain RCS cooldown rate within required TS limits During the initial plant cooldown following a tube leak in the steam generator, the Technical Specification cooldown limit for the reactor coolant system was exceeded. The evaluation of the excessive cooldown determined that there was
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 4 of 36 no adverse impact on the reactor coolant system components and, therefore, is considered a very low risk significant issue. This non-cited violation resulted from the operation crew's deficient monitoring of plant parameters and high pressure steam dump system deficiencies.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to validate and verify an EOP change Deficiencies in emergency operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in a measurable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedure inadequacies Deficiencies in standard operating procedures delayed necessary plant cooldown actions by the operators. The non-cited violation was determined to be an issue of very low risk significance, because the cooldown delay did not result in any appreciable increase in the release of activity during the steam generator failure event.
Inspection Report# : 2000007(pdf)
Mitigating Systems Significance:      Sep 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 6.8 INVOLVING DEFICIENT GUIDANCE IN PROCEDURE AOI 27.1.1 Abnormal Operating Instruction (AOI) 27.1.1, "Loss of Normal Station Power," was deficient, in that no steps were provided in the procedure to identify that the lockout relays for the component cooling water (CCW) pumps were required to be reset following a loss and restoration of power to the motor supply breakers. This deficient procedure is being treated as a Non-Cited Violation of Technical Specification (TS) 6.8, "Procedures and Programs," in accordance with the NRC Enforcement Policy. The consequence of this finding was that the pump lockout relays would have prevented the 21 and 23 CCW pumps from starting automatically on low CCW system header pressure, for 12 days and 21 days, respectively. This finding represented a partial loss of the CCW system function and would reasonably have been corrected by operator action.
Inspection Report# : 2002006(pdf)
Significance:      Aug 10, 2002 Identified By: NRC Item Type: FIN Finding OPERATORS DID NOT IDENTIFY THE APPLICABILITY OF A SHUTDOWN TECHNICAL SPECIFICATION On July 19, 2002, operators did not identify the applicability of a shutdown Technical Specification (TS) associated with the planned removal from service of the 22 emergency diesel generator (EDG) while the 138 kilovolt off site power system was still out-of-service. This finding was associated with the reactor safety cornerstone with respect to
 
3Q/2002 Inspection Findings - Indian Point 2                                                                      Page 5 of 36 the mitigating systems objective of ensuring the availability, reliability and capability of the EDG to respond to initiating events, such as a loss of offsite power, to prevent undesirable consequences. No violation of NRC requirements was identified, since Entergy restored the 22 EDG prior to exceeding the allowed outage time per TS 3.0.1. This finding was of very low safety significance since it did not represent a total loss of emergency power safety function.
Inspection Report# : 2002005(pdf)
Significance: TBD Jul 19, 2002 Identified By: NRC Item Type: AV Apparent Violation APPARANT VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL WILL BE CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING TBD - The team identified an apparent violation of License Condition 2.K of Facility Operating License DPR-26.
License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902. The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC preliminarily risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White).
Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program. This issue is being treated as an apparent violation, consistent with the NRC Enforcement Policy.
(AV 50-247/02-010-001) Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls. At the close of the inspection, Entergy continued to review the design and installation of fire walls particularly, in the areas that interfaced with the central control, cable spreading, and 480 volt switchgear room structures.
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation TURBINE DRIVEN AUX FEED PUMP OIL ISSUES The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning three issues with the control and monitoring of lubrication oil used on the turbine driven auxiliary boiler feed water pump (22 ABFP). Each issue involved incomplete evaluations that led to repeat problems and potential for pump damage. The evaluation and corrective actions following identification in February 2002 that the wrong oil was added to the turbine speed governor were not fully effective. The evaluation of this issue identified that operators were not logging the quantity or specification of oil added during rounds or operation of equipment, but no actions were taken to address the issue. Additionally, the team noted that on July 10, while preparing to run the pump, Entergy identified additional confusion regarding the specification of oil to be added to the governor, an issue that should have been resolved.
Station personnel did not identify that oil analysis results in May 2002 showing a decrease in oil viscosity indicated that the wrong oil was likely added to a pump bearing and that corrective actions for a similar problem previously identified in May 2001 were ineffective. The evaluation and corrective actions following identification in October 2001 of issues with the required oil level in the pump inboard bearing were not fully effective, specifically the design drawing, the vendor manual, and operator training contained inconsistent information. These issues were evaluated using Phase I of the NRC SDP to have very low safety significance (Green), because pump operability was not directly affected. These issues are being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because the issues have been entered into Entergy's CAP. (NCV 50-247/02-
 
3Q/2002 Inspection Findings - Indian Point 2                                                                      Page 6 of 36 010-002 Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SETPOINT DATABASE NOT CORRECTED FOR CIRCUIT BREAKER OVERCURRENT PROTECTION DEVICE SETPOINTS The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to promptly identify, determine the cause, and correct circuit breaker amptector setpoint database errors. The control of design setpoints is necessary to ensure the availability, reliability and capability of safety-related electrical systems. This issue was evaluated using Phase I of the NRC SDP and determined to have very low safety significance (Green), because the team did not identify any instances where a circuit breaker would not have been able to perform its safety function. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP.
(NCV 50-247/02-010-003)
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SAFETY INJECTION TOPPING PUMP VIBRATION CONSEQUENCES TO SAFETY-RELATED PIPING The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to identify that vibration of the non-safety-related SI accumulator topping pump caused stresses in adjacent safety-related piping that were above the code allowable values. The team evaluated this issue using Phase I of the NRC SDP, determining it to have very low safety significance (Green), because liquid penetrant examinations in the areas of high stress did not identify any piping damage. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP. (NCV 50-247/02-010-004 Inspection Report# : 2002010(pdf)
Significance:        Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding MULTIPLE GROUNDS ON THE PROTECTIVE CIRCUIT FOR UNIT 1 SUBSTATION 102NS3 RESULTED IN A LOSS OF THE 13.8 KILOVOLT LIGHTING AND POWER BUS SECTION 3 On May 17, 2002, multiple grounds on the protective circuit for Unit 1 substation 102NS3 resulted in a loss of the 13.8 kilovolt (kv) lighting and power bus section 3. The consequence of this event was a loss of alternate safe shutdown power to all major alternate safe shutdown pumps and selected instrumentation. At the time, the Unit 2 normal and emergency electrical power supplies were available to supply power to the above stated mitigation equipment and instrumentation. The licensee repaired and restored the 13.8 kv bus section 3 within 30 hours of the fault. The performance issue is inadequate retirement of protective circuits for 440 volt substations (132PC3 and 142PC3) that could impact availability of alternate safe shutdown power supplies. This issue is more than minor since unavailability of alternate safe shutdown equipment for 30 hours is viewed as a precursor to a significant event and the alternate safe shutdown power supplies are a risk-significant maintenance rule system which was unavailable for greater than 24 hours.
Inspection Report# : 2002004(pdf)
Significance:        Feb 09, 2002
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 7 of 36 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR A TEMPORARY FACILITY CHANGE INVOLVING THE AUXILIARY FEEDWATER SYSTEM BACKUP NITROGEN SUPPLY SYSTEM.
The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:      Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control M&TE per Appendix B, Criterion XII Entergy identified that measuring and test equipment (M&TE) were out of specification, and that condition reports were not consistently initiated to evaluate the impact of the out of specification M&TE on surveillance tests. Entergy's engineering assessment concluded that the systems impacted by out of specification M&TE were operable. This issue was evaluated in phase 1 of the Significance Determination Process (SDP) and was found to have very low safety significance. A Quality Assurance Audit had previously recognized an inconsistent approach in the control of M&TE.
Although a Business Plan performance improvement initiative exists for this area, progress was insufficient to prevent the observed problems. Contrary to 10 CFR 50 Appendix B criterion XII, the licensee had failed to assure that measuring and test equipment used in activities affecting quality were properly calibrated and adjusted to maintain accuracy within limits. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A. of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:      Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Maintenance per Appendix B, Criterion V The maintenance instructions used to repair the 21 AFW pump on July 16, 2001, were not adequate to pack the pump in accordance with a maintenance standard and vendor instructions. This resulted in poor packing performance and resulted in operators declaring the 21 AFW inoperable during the October 27 shutdown. Further, in 1998 the licensee identified the need to provide instructions on packing pumps to workers, but did not provide adequate information in the maintenance procedures. This issue had a credible impact on safety since a properly packed gland is necessary to ensure reliable AFW pump operation. However, since the maintenance errors did not result in packing failure and a subsequent evaluation concluded the 21 AFW pump could perform its safety function, this issue was determined to have very low safety significance in accordance with a SDP Phase 1 assessment. The failure to provide adequate maintenance instructions for work on safety related equipment was an example of a condition contrary to 10 CFR 50 Appendix B, Criterion V. This violation is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance:      Nov 05, 2001 Identified By: NRC Item Type: FIN Finding CREW HIGH FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 8 of 36 This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with crew high failure rate (four of seven crews failed) during facility-administered annual licensed operator requalification examinations conducted last fall. The finding was previously characterized as having substantial safety significance (Yellow) in NRC Inspection Report 50-247/01-13. The inspectors noted that the licensee's evaluation identified a fundamental underlying weakness: The station has yet to overcome cultural weaknesses that include an unwillingness to confront poor performance, an over reliance on procedures to change behavior, and compartmentalization. More specifically, the licensee identified three root causes: 1) Operations training had not focused on the basic building blocks that ensure a healthy program; 2) The station had not maintained a core of career oriented, plant knowledgeable instructors and operators; and 3) Operations department involvement with Operations Training had often been ineffective. The inspectors concluded that the methodology and level of detail of the licensee's root cause evaluation were reasonable. The licensee implemented a number of corrective actions to address the identified causes. The corrective actions are described in the station's Training Improvement Plan. The more significant corrective actions included initiatives that aimed to 1) improve the quality of training and training materials; 2) increase the number of instructors who have Unit 2 plant experience; and 3) provide additional management support and oversight of training.
The inspectors determined that the corrective actions are appropriately focused on the identified causes. These actions were appropriately prioritized, and either complete or scheduled for completion. Notably, the licensee took strong immediate corrective actions following the requalification examination failures to provide extensive retraining to each shift, and continue to provide this high intensity training. The inspectors independently assessed the extent of the underlying conditions that led to the Yellow finding and found that performance issues had also existed in other Operations Training programs, such as initial licensed operator and non-licensed operator training programs. These problems existed for at least three years, both prior to and following the steam generator tube failure event in 2001.
Although licensee audits and assessments had identified most of the performance problems prior to the crew high failure rate, they did not identify long-term operator performance as a concern. The inspectors concluded that the licensee's extent of condition review appropriately bounded the underlying conditions that led to the Yellow finding as evidenced by the fact that the licensee had also identified the duration and extent of the problems, and the failure to recognize the long standing issues. (Updated) FIN 05000247/01-013-01: Proposed finding due to crew high failure rate during the 2001 annual requalification simulator examinations. This finding was documented in an October 2001 inspection and initially characterized as a potential Yellow finding, the final safety significance to be determined (TBD). This finding was subsequently evaluated under the significance determination process (SDP) and characterized as (reference NRC to Entergy letters dated December 5, 2001, and February 28, 2002). The 95002 Supplemental Inspection (reference Inspection Report No. 50-247/02-09, dated May 31, 2002), assessed the licensee's evaluation of the crew high failure rates and the corrective actions taken to address this performance issue. As stated in the cover letter to Inspection Report No. 50-247/02-09, this finding remains open until after the completion of Entergy's licensed operator requalification examinations, scheduled for September-October 2002, and further review by the NRC. This item remains open.
Inspection Report# : 2002009(pdf)
Inspection Report# : 2001013(pdf)
Inspection Report# : 2002004(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Initial licensee operability evaluation was incomplete-Failure to consider the impact on net positive suction head for the 22 boric acid transfer pump An initial licensee operability evaluation was incomplete in that it failed to consider the impact on net positive suction head (NPSH) for the 22 boric acid transfer pump when the boric acid tank temperature reached 209 degrees Fahrenheit.
This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance:        Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 9 of 36 Failure to consider risk pursuant to 10 CFR 50.65(a)(4)
The licensee failed to fully consider ongoing plant risk with an inoperable main turbine direct trip function between July 21 and August 7, 2001. This issue had a credible impact on safety because of the lack of automatic 6.9 kV bus transfer from the unit auxiliary transformer to the station auxiliary transformer following a postulated 345 kV system fault. On July 22, 2001, the 23 emergency diesel generator was removed from service for planned maintenance. This activity qualitatively would have increased plant risk given a transient on the 345 kV system and short-term unavailability of offsite power to safeguards buses 2A and 3A with no emergency power to safeguards bus 6A during the planned maintenance. Operator actions would be necessary to restore power to two of four safeguards buses.
Qualitative assessments were not performed until the inspector discussed this observation with the licensee on August 7, 2001. Additionally, risk associated with the inoperable trip should have been incorporated into maintenance restrictions on certain safety equipment. This issue was evaluated in the Significance Determination Process and found to have very low safety significance. The failure to consider plant risk for an inoperable main turbine direct trip from a 345 kV fault is contrary to 10 CFR 50.65(a)(4). This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368)
Inspection Report# : 2001008(pdf)
Significance:      Jul 07, 2001 Identified By: NRC Item Type: FIN Finding GAS TURBINE 2 FOUND TO BE INOPERABLE DURING ROUTINE MONTHLY TESTING Gas Turbine 2 was found to be inoperable during routine monthly testing on May 28, 2001. GT-2 remained out of service for eight days as Con Edison continued to identify and investigate several support system problems. The problems and degraded material conditions were long-standing and were present despite the recent extended maintenance outage to overhaul GT-2. The untimely resolution of long-standing degraded conditions was a contributor to an adverse performance trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance:      Jul 07, 2001 Identified By: Self Disclosing Item Type: FIN Finding Failure of a fire water header.
During a test of the fire water system on June 5, 2001, the 12 inch fire water header failed, which resulted in a leak of 231,000 gallons of city water into the Utility Tunnel. The automatic and manual fire suppression system was inoperable for approximately 1 hour and 15 minutes, which impacted 14 fire zones that contained alternate safe shutdown equipment. The licensee restored the main fire header back to a fully functional status on June 10, 2001. The fire header failed because of inadequate alignment and torque setting of the Victaulic couplings when the header was modified in November 2000. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001006(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: FIN Finding MAINTENANCE RISK ASSESSMENT AND EMERGENT WORK Gas turbine #1 (GT-1) failed during a test on May 3, 2000. Con Edison identified degradation in the turbine and compressor sections, and noted significant cracking in the first stage stationary blades. A preliminary assessment concluded the degradation was significant and questioned whether GT-1 could have operated for its design basis mission time. The plant risk associated with all three gas turbines potentially inoperable for a 24 hour period in March 2001 was reviewed using the Significance Determination Process and had a very low safety significance. GT-1 remained out of service pending disassembly, inspection, repair assessment, and a formal operability assessment.
Inspection Report# : 2001004(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 10 of 36 Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN ADEQUATE RECORDS OF REQUALIFICATION ATTENDANCE Con Edison did not have attendance records for an average of 30% of the licensed operator training classes for the years 1998-2000. This issue has minimal safety significance because the facility was able to provide examination/evaluation records of program participation. Con Edison verified operator attendance through written and simulator evaluation records. Corrective actions were addressed in Condition Report 200008293. The failure to have complete records of licensed operator training was contrary to the 10 CFR 55.59(c)(5) and the record retention requirements of Technical Specification 6.19.2.g. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance: N/A May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE POST MAINTENANCE TESTING Con Edison identified that corrective actions were not effective to correct a violation related to the completion of post-maintenance testing (PMTs). There were no operability or safety issues related to the outstanding PMTs for safety related equipment that had been returned to service. This matter was a repetitive, licensee-identified violation of TS 6.8.1 having minimal safety significance for the failure to have documented assessment of the outstanding PMTs. This item is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ADDRESS THE EFFECT OF AMBIENT TEMPERATURE ON THE SETPOINT OF MAIN STEAM CODE SAFETY VALVES The NRC identified that Indian Point Unit 2 failed to take adequate corrective actions to address the effect of ambient temperature on the setpoint of main steam code safety valves, in response to a prior NRC violation, related to pressurizer code safety valve setpoint testing. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:      May 19, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE MAIN STEAM CODE SAFETY TESTING WAS ADEQUATE WHILE USING A LIFT ASSIST DEVICE The NRC identified that Indian Point Unit 2 (IP2) failed to establish measures to ensure that main steam code safety testing requirements were implemented, while making use of a lift assist device. Because there was no indication that an actual loss of safety function occurred, the Significance Determination Process screened this condition as one of very low safety significance. This violation of IP2 technical specification 4.2.1, Inservice Testing, has been entered in Con Ed's corrective action system and is being treated as a non-cited violation.
Inspection Report# : 2001004(pdf)
Significance:      Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item
 
3Q/2002 Inspection Findings - Indian Point 2                                                                    Page 11 of 36 Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to have adequate length of fire hose staged for manual fire fighting in the central control room The team determined that the 100 feet long fire hoses on the primary and secondary hose reels for central control room (CCR) were too short to reach all areas of the CCR. ConEd took immediate corrective action to stage additional hose lengths near the primary hose station for the CCR, and documented the deficiency in the corrective action program. The failure to be able to reach all areas of the CCR with 100 feet length fire hose is a violation of the Fire Protection Program Plan, which is incorporated into the operating license, by reference, in License Condition 2.K. The significance determination process characterized this condition as being of very low risk significance because the control room is continuously manned, and most fires would be detected and extinguished at the incipient stage using portable extinguishers. This violation of the operating license is being treated as a Non-Cited Violation (NCV 050000247/2000-004-02), consistent with Section VI.A. of the Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate isolation of circuits from the central control room The team found that the remote control switches, and their associated wiring, in Unit 1 control panel board located in the CCR of several 13.8 kV light and power breakers (SB1-2, SB1-3, SB1-T, SB2-2 and GT-1) of Alternate Safe Shutdown System (ASSS) power supply were not capable of being isolated from central control room circuit wiring, an area for which the system is credited. This is contrary to section III G.3 of Appendix R. In the event of a fire in the control room, the control of these breakers could be adversely affected and the alternate safe shutdown power relied upon could become unavailable. No procedural steps exist to recover these breaker functions. ConEd entered this deficiency into the corrective action program on April 13, 2001, to address this issue. The team determined that this issue was of very low risk significance (Green). This violation of 10 CFR 50, Appendix R, section III.G.3 requirement, not providing adequate isolation of circuits from the central control room, is being treated as a non-cited violation (NCV 050000247/2000-004-03), consistent with Section VI.A. of Enforcement Policy.
Inspection Report# : 2000004(pdf)
Significance: N/A Apr 13, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate Document Control for RPS Wire Lists 10 CFR 50, Appendix B, Criterion VI, "Document Control," requires measures to be established to control the issuance of documents, such as instruction and drawings, including changes thereto. Con Edison did not adequately control the issuance of the RPS wire lists (controlled documents) in that the errors referenced in CR 200008415 (annunciator circuits incorrectly listed in reactor trip listing, incorrect relay numbers and incorrect relay locations) were not corrected. In addition, the RPS wire lists had not been properly updated to incorporate the wiring changes for the P-10 relay contacts in 1982, and the relay replacement/modification in December 2000. The corrective actions for this violation were already in Con Edison's corrective action program. This is a non-cited violation.
Inspection Report# : 2001005(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 12 of 36 Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for EDG Maintenance During preventive maintenance on the 22 emergency diesel generator (EDG) in March 2001 per ICPM 1780, a technician identified an incorrect configuration on the fuel oil primary filter differential pressure switch for all three emergency diesel generators. Procedure ICPM 1780 did not provide sufficient guidance to detect the configuration problem when the same calibration was performed in 1998 and 1999. This issue did not result in a loss of diesel generator function and had very low safety significance. The failure to provide adequate procedures for EDG maintenance was a Non-Cited Violation of Technical Specification 6.8.1.a. NCV 2001-003-01 Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply The operators identified a failed status light on the train A blackout without safety injection logic circuit, but failed to complete a timely evaluation per AOI 10.1.4 to identify that a blown fuse had de-energized the power supply. This resulted in the untimely detection of a loss of redundancy in the engineered safety features logic. Since the failure did not result in a loss of safety function and the plant was operated within the Technical Specification Table 3.5-3 limiting condition of operation, this issue had very low safety significance. Other performance issues noted included incomplete information provided in the shift turnover brief, the lack of clear guidance in the procedures used to diagnose circuit problems, and the lack of clear directions in the technical specifications on implementing the limiting condition for operation.
Inspection Report# : 2001003(pdf)
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Tagging Controls - CST Inventory Loss The failure to control tagged equipment resulted in a diversion of approximately 20,000 gallons of inventory from the condensate storage tank, which is the inventory source for the secondary heat removal system. Operations Administrative Directive (OAD)-36 requires that workers inform the control room if operation of a component with a caution tag is desired. Contrary to OAD-36, security personnel inadvertently manipulated a temporary breaker that was caution tagged without informing the operations crew. The event did not result in a loss of safety function and the TS limiting condition of operation for the condensate storage tank was not exceeded. This issue had very low safety significance. This violation is being treated as a Non-Cited violation of Technical Specification 6.8.1.a. This is an example of a configuration control problem.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding GT-3 found inoperable while GT-2 was out of service for corrective maintenance and EDG 22 were out of service for planned maintenance During an extended outage on gas turbine 2 (GT-2) for corrective maintenance and a planned outage on EDG 22 for preventive maintenance, GT-3 became inoperable due to loss of air pressure, as indicated by an alarm and lock-out
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 13 of 36 from pressure switch PS-11. The low pressure lock-out occurred when workers used the GT-3 air system to run air-operated tools for the work on GT-2, and could not be cleared initially when the air service was returned to normal.
Followup investigations determined that PS-11 was functioning properly, but the pressure lock-out needed to be reset manually, and that requirement was neither known by the operators nor covered in the procedure. Although GT-1 remained operable to satisfy the TS 3.7.C.1 requirements, the loss of GT-3 caused the plant daily risk factor to increase from 2.01 to 5.44 for about 23 hours. This issue had very low safety significance.
Inspection Report# : 2001003(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions - 22 ABFWP oil loss The 22 auxiliary boiler feedwater pump (ABFWP) became inoperable when workers accidently opened a drain valve which caused the loss of oil in the outboard bearing. While actions were taken to identify the adverse condition, assess the pump condition and restore it to an operable status in a timely manner, the followup corrective actions did not address actions to prevent recurrence until questioned by the NRC. The event did not result in the loss of the secondary cooling system safety function and the 22 ABFWP was inoperable less than the TS allowed outage time. Therefore, the specific issue had very low safety significance. However, the inoperability of this risk-significant pump is of concern.
For example, an NCV was issued in NRC Inspection 05000247/2000-12 for the failure to implement corrective actions to prevent recurrence for the inadvertent operation of the 22 ABFWP overspeed trip device. NCV 2001-003-03 Inspection Report# : 2001003(pdf)
Significance:      Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow surveillance procedures With the plant operating at 100% full power on February 14, 2001, power was lost to 480 volt Bus 3A during a test of safety bus undervoltage relays. The event was caused by technician error in failing to follow the test procedure. This issue had low safety significance because the loss of safety Bus 3A was of short duration and the remaining multi-train systems were available. The failure to follow procedures was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance:      Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow operating procedures On January 2, 2001, with the unit at 6.5% full power, a main turbine trip signal was generated by a high level in the 21 steam generator. The high steam generator level tripped the main boiler feed pump and actuated the auxiliary feedwater system. Three operator or crew performance problems were identified and consisted of the following: the failure to adequately control steam generator level; operator control of rod insertion without a complete understanding of reactor conditions; and, operator communication errors, which resulted in an unnecessary plant cooldown and the simultaneous insertion of reactivity by two means. The issue was evaluated using the NRC's significance determination process as having low safety and risk significance. The failure to operate the reactor in accordance with procedures for reactivity management and controlling reactor temperature was a non-cited violation of Technical Specification 6.8.1.a.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: FIN Finding Findings of a number of human performance issues The inspection findings this period, and other issues documented in the corrective action process, indicated a number of
 
3Q/2002 Inspection Findings - Indian Point 2                                                                    Page 14 of 36 human performance issues, some of which had significance relative to personnel safety, plant operation or plant equipment. NRC concerns with the number and significance of human performance errors were discussed with the Plant Manager in a meeting on February 16, 2001. The licensee described actions and plans to address this issue.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to make timely notifications Review of the January 2 event to evaluate performance and procedure adherence was hampered by poor log-keeping practices, untimely and undocumented operator interview information, and poor plant data retrievability. The initial management response to the event was incomplete and allowed power escalation to continue with incomplete short term actions outstanding. The initial licensee reviews did not identity the procedure adherence and reactivity control issues. Subsequent review by the event review team identified that startup pressures potentially impacted operating activities. Followup actions to address this concern were appropriate.
Inspection Report# : 2000015(pdf)
Significance: N/A Feb 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow log keeping procedures The failure to implement procedure requirements for log keeping was a non-cited violation of Technical Specification 6.8.1.a. The log keeping violation was considered more than minor because corrective actions from August 31, 1999, and February 15, 2000, events were not completely effective. The failure to make timely notification to the NRC of an actuation of the auxiliary feedwater system was a non-cited violation of 10 CFR 50.72(b)(2).
Inspection Report# : 2000015(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50 Appendix B, Criteria XVI, Corrective Action The licensee failed to identify and correct the cause of repetitive failures of the service water strainers and motor operated service water isolation valve SWN-7. These items were determined to be of very low safety significance because the strainer failures did not have more than a minimal impact on system operability and the valve failures were identified when the valve was out of service for maintenance.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports - service water strainer blowdown flow rates The licensee failed to initiate condition reports for three failures to meet the acceptance criteria for service water strainer blowdown flow rates during the performance of procedure PT-93 on July 13, 2000. This issue was determined to be of very low safety significance because the operability of the system was not affected.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Building Ventilation System The design termperature ratings of electrical components in the emergency diesel generator (EDG) building, including
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 15 of 36 ventilation fan thermal overloads, cabling, and control power transfer switches had not been verified. These issues were of very low significance because the as-found thermal overload settings would not have resulted in the loss of ventilation at the maximum building temperatures, the effects of elevated temperature on the cabling voltage drop calculation would have been negligible, and information obtained from the vendor indicated that the control power transfer switch circuitry would have remained functional at the elevated temperature.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation EDG Manual Load Control The results of the EDG loading calculation had not been transmitted to the operations department for inclusion into appropriate operating and test procedures. These issues were of very low safety significance since the ability of the EDGs to provide emergency power was not affected and the procedure issues would not have impacted safe operation of the affected systems.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Header Pressure Analyses The ability of the service water system to supply adequate flow to all safety-related components based on existing service water low header pressure alarm setpoint and the control room log limits was not supported by engineering calculations. The licensee performed a preliminary analysis and detrmined that the alarm setpoint of 53 psig was adequate to ensure adequate flows. However, if pressure decreased to the control room log limit of 48 psig the system would not have had sufficient capacity to supply adequate flow to all components. The licensee increased the control room log limit to 58 psig, giving a 5 psig margin to the 53 psig low pressure alarm design limit. This issue was of very low safety significance because there was no indication that the service water system had been operated below a header pressure of 53 psig.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Flooding Controls were not in place to prevent damage to components in the service water strainer room given an external flood caused by high river water level and a concurrent internal flood due to a potential single failure of a service water pump vacuum breaker valve. The licensee implemented a temporary procedure change to address this issue. This issue was of very low safety significance because it involved the relatively low probability of an internal flooding event coupled with the low probability of an external flooding event.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Temporary Procedure Change Process Appendum VI to SAO 100, "Indian Point Station Procedure Policy," Rev. 3, which describes the process for implementing temporary procedure changes (TPCs), was not followed when alarm response procedure ARP AS-1 (Accident Assessment Panel 1; windows 5-4 and 6-4) was changed with TPC 00-0853, This TPC was implemented
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 16 of 36 because a temporary modification disabled the associated alarm inputs; however, the alarm inputs had already been disabled and the change was not required for immediate operation of the plant. This issue was of very low safety significance because the use of a TPC did not have any actual detrimental affect on plant operations.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Control of Setpoints for Delta - Temperature Annunciation The reactor coolant loop Delta-Temperature alarm was received during power ascension as a result of having an incorrect setpoint value in calibration procedure. This issue was determined to be of very low safety signficance since the instrument does not have any automatic protective function, only an alarm function.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Oil Pads in EDG Instrumentation Cabinet Leaving two oil absorbent pads inside the EDG 21 instrumentation cabinet following repairs to a leak did not comply with SAO-701, "Control of Combustibles and Transient Fire Load," Rev. 8. This issue was of very low safety significance because it did not represent a fire impairment nor a degradation of a fire protection feature or defense in depth issue.
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Service Water Strainer Pit Drain Check Valve The plant testing program did not include a verification that the safety-related service water strainer room drain line check valve, MD-500, could open to prevent internal strainer pit flooding. The licensee demonstrated operability by manually cycling the valve from the full open to full closed position and observing that the valve opened with minimal effort and that there was no restriction in movement. This failure to test a valve by periodically exercising it to its safety function position is being treated as a non-cited violation of 10 CFR 50.55a, "Codes and Standards," paragraph (f),
"Inservice Testing Requirements."
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Gas Turbine Performance Corrective actions were not taken to resolve reliability and availability performance issues with the alternate AC power sources, gas turbines (GTs) - 1, -2 and -3. The GTs had not been meeting the licensee developed maintenance rule reliability and availability performance goals since 1995. The team did an independent calculation of the change in core damage probability associated with te unavailability of GT-2 for an estimated repair length of 60 days and determined the risk increase to be within the very low safety significance band (<1E-6). This issue was of very low safety significance because the Technical Specifications relative to GT availability were met. This failure to effectively implement corrective actions to ensure that the established maintenance rule goals would be met is being treated as a non-cited violation of 10 CFRR 50.65 (a)(1).
Inspection Report# : 2001002(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 17 of 36 Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Systems not Operated As Designed Design bases information was not translated into electrical systems testing and operating procedures acceptance criteria or operating limits. This issue was of very low safety significance because none of the test results or operating data reviews identified instances where equipment was operating outside of its design limits. This failure to include appropriate acceptance in the procedures and drawings to ensure activities have been satisfactorily accomplished is being treated as a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2001002(pdf)
Significance:      Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Fuel Oil Transfer Procedure Abnormal Opersating Instruction (AOI) 27.3.1., "Emergency Fuel Oil Transfer Using the Trailer," Rev. 0, did not provide adequate instructions for filling the trialer. This issue was of very low safety significance because the use of this procedure has never been required and would require minor changes to resolve the discrepancies.
Inspection Report# : 2001002(pdf)
Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate the design basis into procedures The licensee did not have a formal process for implementing changes to the plant licensing basis, and certain limits and provisions of two technical specification amendments were not adequately incorporated into plant operating procedures. As a result, there was the potential to have exceeded the technical specification analytical limits on safety injection accumulator pressure, and post-accident radiological doses to control room operators could have exceeded analyzed limits. The conditions had a potential impact on safety in that fuel peak cladding temperature and control room habitability could have been adversely affected. If left uncorrected, inadequate implementation of license amendments could result in a more significant safety concern. The conditions were evaluated using the NRC's significance determination process as having very low safety significance because no actual loss of safety function occurred. This violation of the design control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: URI Unresolved item Evaluation of RWST Design The team noted that a formal calculation is pending for the deliverable volume from the RWST that accounts for level instrument uncertainties. The NRC raised questions on the available tank vent area; seismic adequacy of overflow line, and criteria for securing containment spray pumps. These issues would not impact system operability. An open item will track the completion of these evaluations and NRC review.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct inadequate design interfaces
 
3Q/2002 Inspection Findings - Indian Point 2                                                                    Page 18 of 36 No Color - The NRC identified that the lack of formal design interface controls that are required by Criterion III of 10 CFR 50, Appendix B, and the licensee's Quality Assurance Program Description had been identified previously by the licensee's Quality Assurance organization and the NRC. Failure to promptly correct this condition adverse to quality resulted in multiple discrepancies between design inputs used in accident analyses and actual plant conditions and procedures. The matter had a potential impact on safety due to the potential effects on safety margins, which left uncorrected could become a more significant safety concern. This issue had a very low safety significance because the design discrepancies involved did not result in the actual loss of safety function. This violation of the corrective action requirements of 10 CFR 50, Appendix B, Criterion XVI was treated as a non-cited violation consistent with Section VI.1.A of the Enforcement Policy due to the very low safety significance of the specific design discrepancies involved.
Inspection Report# : 2000014(pdf)
Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish measures for control of design interfaces The licensee does not have formal procedures to control the verification, validation, and supply of input data and assumptions to the NSSS vendor, and administrative controls were not adequate to ensure that accident analysis input assumptions were not invalidated by plant modifications. As a result, discrepancies existed between the values assumed in certain accident analyses and actual plant conditions and procedure limits. The discrepancies had potential adverse impact on post-accident fuel peak cladding temperature and containment peak pressure. If left uncorrected, the lack of formal control of design inputs could become a more significant safety concern. The specific conditions caused by the lack of formal design controls were evaluated using the NRC's significance determination process as having very low safety significance because of the limited actual consequences of the input discrepancies on the accident analysis conclusions, and no loss of safety function occurred. This violation of the design interface control requirements of 10 CFR 50, Appendix B, Criterion III was treated as a non-cited violation.
Inspection Report# : 2000014(pdf)
Significance:      Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take corrective actions for freeze protection Corrective actions were ineffective to prevent recurrence of material condition concerns with the freeze protection for the refueling water storage tank (RWST), primary water storage tank (PWST) and condensate storage tank (CST) level switches. Over the last three years several condition reports associated with the material condition of the freeze protection for these level switches had been generated, some of which were associated with actual failures of the switches. Although in each case corrective actions were taken to address the specific failure, no corrective actions were taken to prevent recurrence of problems with the freeze protection of these level instruments. This issue had a very low safety significance because it did not result in the actual loss of a safety function. The failure to take corrective actions to preclude repetition is being treated as a non-cited violation of 10CFR50, Appendix B, Criterion XVI, "Corrective Action."
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding The NRC evaluated Con Edison's actions to review plant systems prior to restart The NRC evaluated Con Edison's actions to review plant systems prior to restart. No operability issues were identified during system walkdowns and status reviews. Most deficiencies were identified by Con Ed; one exception was a problem with a safety injection system pipe support. The NRC noted mixed quality with some walkdowns because system engineer preparation appeared inconsistent and some knowledge weaknesses were noted. Some improvements and procedure changes were made, and some systems were reviewed again. Management review of system health presentations met the intent of the administrative procedures. The initial reviews did not appear to be particularly
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 19 of 36 probing of the conclusions on system health; improvements were noted in later presentations. NRC review of system health continued at the conclusion of the inspection.
Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Thermal Sleeve Con Edison completed action to evaluate a degraded thermal sleeve in the #23 cold leg pipe of the reactor coolant system (RCS) and retrieved loose pieces. The licensee had previously evaluated the thermal sleeves using radiography earlier in the 2000 refueling outage and incorrectly concluded that #23 was intact. The findings this period revealed that the radiographs had been incorrectly interpreted. Con Edison completed a foreign object search and retrieval (FOSAR) after the lower internals were removed and recovered the remnants of the #23 thermal sleeve. Con Edison determined that IP2 can safely operate without a thermal sleeve and with any remaining piece(s) in the RCS Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding 23 Auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker The 23 auxiliary feedwater pump failed to start during a surveillance due to an electrical problem with the DB-50 supply breaker. The specific failure had low safety significance because the breaker that failed was installed during the present outage. Corrective actions considered the extent of condition for other DB-50 breakers. This appears to be a missed opportunity for the corrective action and preventive maintenance programs to have identified high contact resistance in the breaker closing circuit prior to a demand failure of a safety related component Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Following replacement of Battery Bank 22, the battery failed a modified performance test Following replacement of Battery Bank 22, the battery failed a modified performance test when the capacity dropped below 90% (89.7%) prior to the end of the 4 hour test interval. The battery was installed while the plant was shutdown.
The battery was considered functional because the capacity was greater than the design basis requirement to provide essential loads for two hours. However, the 22 Battery failed a capacity test on three previous tests during the present outage. Con Edison reported this matter to the NRC per 10 CFR Part 21 by {{letter dated|date=November 16, 2000|text=letter dated November 16, 2000}}, based on a potential defect in the manufacture of the cell plate material. Batteries 21, 23 and 24 have operated and tested satisfactorily. Con Edison continued to evaluate the battery performance and prepare an operability determination Inspection Report# : 2000013(pdf)
Significance:      Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Maintenance Risk Assessments and Emergency Work Control Con Edison implemented Modification FPX-00-12449-F to address degraded relay conditions and eliminate a potential for multiple relay failures. The reactor protection system (RPS) was not required to be operable since the work was done while the reactor was in cold shutdown. Although the relays had remained functional, the replacement was deemed appropriate to assure the debris from degraded coils would not prevent proper relay operation. The inspector
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 20 of 36 verified that the combination of work controls and post-work testing would provide assurance that the RPS would be operable for subsequent plant operations.
Inspection Report# : 2000013(pdf)
Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Steam generator replacement project The activities of the IP2 steam generator replacement project (SGRP), including transport and storage of steam generators, the eddy current inspection of tubes in the replacement steam generators, in-progress radiography of welds, provision for reinstallation of components removed as part of the SGRP and control of work package closeout were noted to be well planned and conducted. Radiation surveys for interim storage of the old steam generators showed measured radiation levels to be below regulatory limits.
Inspection Report# : 2000013(pdf)
Significance:        Nov 18, 2000 Identified By: NRC Item Type: FIN Finding Utility Tunnel - Unit 2 support services Con Edison completed a risk significance evaluation of the components in the Utility Tunnel. The evaluation consisted of a functionality assessment of the mechanical and electrical components in the tunnel that were degraded due to inadequate supports and pipes corroded from ground water ingress into the tunnel. Portions of the fire protection header were replaced this period to address areas of severe wall loss. Long term corrective actions remained in progress to conduct additional engineering walkdowns to identify abandoned services that should be removed as a modification, and finalize long term repairs and upgrades.
Inspection Report# : 2000013(pdf)
Significance:        Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate fire fighting strategy instruction existed to align fire suppression water to containment An inadequate fire fighting instruction existed to align fire suppression water to the containment. The deficiency impacted the efforts to suppress the fire inside containment on September 3, 2000. This issue had very low risk significance because safe shutdown equipment was not impacted by the fire. A violation of license condition 2.K is being treated as a non-cited violation Inspection Report# : 2000011(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: FIN Finding Damaged Service Water Pump and Motor Control Center 21 Power Cables Con Edison identified damage to the power cables for motor control center (MCC) 21, service water pumps (SWPs) 25 and 26, and feeds for other non-essential intake loads. The cables were damaged when a duct bank routing cables to MCC-21 settled at the intake structure The SWPs remained functional up to the time the condition was discovered and were removed from service while repairs were completed. The other four service water pumps were not affected. The licensee's preliminary evaluation of the condition included a root cause evaluation and provided the bases for a conclusion that the service pumps remained operable under assumed seismic conditions. Civil repairs and modifications were completed, and the affected MCC-21 and service water pump cables were replaced. The condition occurred due to a combination of stresses applied to the duct bank when the original cables were installed, and inadequate support for the duct bank at the intake foundation. The licensee planned to continue investigations of the soils in the intake area. The licensee entered this issue in the corrective action program as Condition Reports
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 21 of 36 200003630 and 200004004. The risk associated with the degradation of the service water pump cables was reviewed by the regional senior Reactor Analyst. This condition would be a very low risk condition (GREEN). This is based on the fact that the cables had not failed and the safety function would likely have been performed.
Inspection Report# : 2000008(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly document and accept the bases for the OD The final calculation for the charging pump seal water tank, which provided the long term basis for operability, was not approved, accepted or entered into the Con Ed Calculation Indexing Program contrary to procedure requirements. This issue was determined to have very low risk significance since the equipment operability was not impacted. Deficient control, review and approval of these calculations and of the associated operability determination are collectively considered a violation of 10 CFR 50, App. B, Criterion V and is being treated as an NCV.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly identify and evaluate the full scope of the modification in the SE The safety evaluation for a modification to the chemical volume and control system power supply did not completely define the scope of work. The safety evaluation incorrectly stated that the associated modification did not add any new wires or cables. The failure to assess the full scope of the modification in the safety evaluation was determined to be a non-cited violation. Failure to include and evaluate the new cables in the safety evaluation was determined to have very low risk significance because it dide not change the overall conclusions reached in the safety evaluation regarding an unreviewed safety question, and did not adversely impact the plant design modification.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct deficiencies associated with the steam generator nitrogen 16 monitors Con Edison did not take timely corrective actions for the steam generator leak monitoring recorder deficiency. The failure to take adequate corrective actions was determined to be a non-cited violation and was an issue of very low risk significance in that there was a minimal impact on the operators' ability to determine the magnitude of the steam generator tube leak.
Inspection Report# : 2000007(pdf)
Significance:      May 20, 2000 Identified By: NRC Item Type: FIN Finding The licensee identified a degradation in thye boraflex panels in the spent fuel storage racks The licensee identified a degradation in the boraflex panels in the spent fuel storage racks, which resulted in a plant condition outside the design basis. Con Edison monitored degradation in boraflex panels in spent fuel pool racks using surveillance coupons, pool chemical analyses and analytical simulations using a computer program. On April 6, 2000, the results of boron-10 areal density measurements showed that thinning had occurred and gaps up to 7 inches had formed in the boraflex panels. Conservative criticality analyses assuming worst case gap size and geometry showed that the design requirement established in the technical specifications could not be met. Technical specification (TS) 5.4.2.B requires that the storage racks be designed such that the effective multiplication factor (Keff) is less than 0.95
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 22 of 36 without soluble boron in the pool water. The NRC Safety Evaluation for License Amendment No. 158 described the use of administrative controls such as fuel assembly relocation to compensate for boraflex degradation. Con Edison used additional controls on soluble poison concentration and spent fuel loading patterns to assure the Keff requirements were satisfied. This issue was considered to have a very low risk significance (Green) using the Significance Determination Process (SDP) phase 3 evaluation, because the storage rack Keff remained below 0.95 during past periods when a checkerboard pattern was not used but soluble boron concentration was at least 1500 ppm.
Inspection Report# : 2000005(pdf)
Significance:      May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Manipulator Crane The licensee failed to maintain adequate control of the manipulator crane control circuits. The circuit wiring was not in accordance with controlled drawings. A jumper bypassed a safety feature in the manipulator crane control circuit. With the jumper installed, the manipulator crane gripper could have been released prior to the fuel assembly being fully lowered into the core. The manipulator crane load cell interlock was not affected. The circuit would have prevented the operator from releasing the gripper under load and dropping a fuel assembly. The event was reviewed with the regional Senior Reactor Analyst (SRA), who evaluated the safety significance as very low (Green) based on the fact that the load cell remained operable and the procedural requirement for the operator to verify the location of the fuel assembly prior to releasing the gripper. The failure to maintain adequate design controls was determined to be a non-cited violation of 10 CFR 50, Appendix B, Criterion III. This inadequate control did not have an actual impact on safety.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:      Jun 29, 2002 Identified By: NRC Item Type: FIN Finding DURING SURVEILLANCE TESTING OF THE SAFETY INJECTION DISCHARGE MOTOR-OPERATED VALVE 851B, THE VALVE FAILED TO STROKE CLOSED On May 27, 2002, during surveillance testing of the safety injection discharge motor-operated valve (851B), the valve failed to stroke closed. The initial operability evaluation did not consider the non-automatic containment isolation function for this valve. This event was documented in condition report No. 200205433. The performance issue associated with this finding is a weakness in operator knowledge of multi-function safety system components. This is the second recent example where operators did not consider this function for a safety-related valve. The first example was documented in NRC report 50-247/2002-003, section 1R15. The untimely and incomplete operability assessment for safety injection discharge valve 851B has very low safety significance since the containment isolation valve was restored to an operable status prior to exceeding Technical Specification 3.6.A.3.a.2.d limiting condition for operation.
Inspection Report# : 2002004(pdf)
Significance:      May 11, 2002 Identified By: Licensee Item Type: FIN Finding UNTIMELY OPERATOR EVALUATION FOR CONTAINMENT ISOLATION VALVE 869B On April 11, 2002, operators did not complete a timely operability evaluation for containment isolation valve 869B after the disconnect switch operating handle on motor control center (MCC)26BB broke while applying an equipment tagout. At the time, the operators neither verified that the disconnect would operate nor completed an adequate evaluation regarding the ability to close valve 869B to perform its containment isolation function. An operability
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 23 of 36 evaluation was completed about six hours later by a different operating crew and the operators then entered a four-hour limiting condition for operation and isolated the containment penetration per the technical specifications 3.6.A.3.a.2.b.
The untimely operability evaluation increased the unavailability time for the containment spray system. The inoperable containment isolation valve issue was more than minor because it impacts the containment barrier. This issue had very low safety significance since the containment isolation valve was repaired and restored to an operable status prior to exceeding technical specification 3.6.A.3.a.2.d. This issue was an example of untimely operator implementation of technical specification requirements in response to degraded safety equipment.
Inspection Report# : 2002003(pdf)
Significance:      Feb 09, 2002 Identified By: Licensee Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release. The failure to maintain containment integrity was a violation of TS 3.6.
This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:      Dec 29, 2001 Identified By: Licensee Item Type: NCV NonCited Violation MULTIPLE FAILURES TO ADHERE TO TS FIGURE 3.1.4-2 DUE TO INADEQUATE PROCEDURES IN THE YEAR 2000 Entergy determined that the plant operated in violation of the RCS overpressure protection requirement of TS Figure 3.1.A-2 on four separate time periods in the year 2000 with a total exposure of approximately 49 hours. The cause was the failure to account for instrument errors in operating procedures used for controlling plant conditions in accordance with TS Figure 3.1.A-2. This issue was evaluated in the SDP process (Manual Chapter 0609 Appendix G) for a violation of the low temperature overpressure protection technical specifications. During the times when the facility operated outside TS Figure 3.1.A-2, all appropriate administrative controls to limit the potential for unwarranted heat-up or mass addition to the reactor coolant system were implemented by operators. The consequence of this error potentially reduced the required operator response time for a postulated overpressure events as previously approved in the plant licensing basis. No reactor coolant system overpressure condition existed during these times and the 10 CFR 50 Appendix G limits were not exceeded. However, the multiple failures to adhere to TS Figure 3.1.A-2 due to inadequate procedures is considered a violation of TS 3.1.A.4 and TS 6.8.1.a. These violations are treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368). A TS Amendment was submitted and was under review at the end of the inspection.
Inspection Report# : 2001011(pdf)
Significance:      Jul 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified steam generator eddy current inspection technique for U-bend areas during the 1997 outage During the 1997 refueling outage the U-bend mid-range Plus Point ECT probe, used for SG tube inspection, was not
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 24 of 36 properly set up to the correct calibration standard. Specification NPE-72217 required the use of an Electric Power Research Institute (EPRI)-qualified technique. The probe was not set up with the calibration standard or with the phase rotation specified on the EPRI qualified technique #96511, dated May 1996. This issue did not have a substantial impact on the ability to detect PWSCC flaws. This issue involved matters with very low risk significance, because it did not directly affect the ability to detect tube flaws and as such, did not affect the reactor coolant system integrity.
The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Special Processes.
Inspection Report# : 2000010(pdf)
Significance: N/A Jul 20, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator program ineffective corrective actions during 1997 outage The team concluded that Con Edison's root cause analysis for the SGTF, dated April 14, 2000, did not identify and address significant SG inspection program performance issues as they related to the failure of tube R2C5 in SG 24 on February 15, 2000. While the root cause analysis attributed the SGTF to a flaw that was obscured by ECP signal noise, it did not identify or address deficiencies in the processes and practices during the 1997 SG inspection.
Inspection Report# : 2000010(pdf)
Emergency Preparedness Significance:        Sep 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECT PREVIOUSLY IDENTIFIED CONDITION IN THE JNC REGARDING THE TIMELY AND ACCURATE DISSEMINATION OF INFORMATION During the emergency plan exercise conducted on September 24, 2002, the licensee JNC personnel proceeded with a 1:55 p.m. press conference informing the media that no release was in progress when, according to the exercise events, a release had begun just prior to the press conference. This information was not in conflict with the emergency alert system message that was in effect at the time of the briefing. However, the failure to provide updated information could cause confusion for those receiving it through the media. This failure was a previously identified weakness as documented in Drill Critique Reports and in condition reports. This is a non-cited violation of 10 CFR 50 Appendix E Section IV.F.2.g which requires, in part, that weaknesses or deficiencies that are identified during a drill or exercise shall be corrected. The risk associated with this release of incorrect information was determined to be of very low significance because it does not constitute a loss of function in meeting the applicable planning standard (10 CFR 50.47 (b)(14).
Inspection Report# : 2002012(pdf)
Significance: N/A May 11, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF 10 CFR 50.54(q) FOR ACCOUNTABILITY On March 6, 2002, the licensee implemented changes to the accountability process that decreased the effectiveness of the Emergency Plan (E-Plan). The finding was considered more than minor because, if left uncorrected, it would become a more significant safety concern. Changing commitments in the E-Plan without prior approval impacts the NRC's ability to perform its regulatory function and potentially creates an ineffective response to a radiological emergency. The consequences of this change were minimal because it did not preclude the function of accountability from being performed, albeit delayed. The licensee has implemented the corrective actions and has since met the timeliness goal. This change which decreased the effectiveness of the Plan is being treateed as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued May 1, 2000.
Inspection Report# : 2002003(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 25 of 36 Significance:        Jun 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct a bi-weekly silent test as specified in the licensee's emergency plan A non-cited violation of 10 CFR 50.54(q) was identified. Licensees are to maintain and follow their emergency plan.
The NRC determined that the licensee did not conduct a bi-weekly silent test within the required periodicity as specified in Section 6.6 of the emergency plan during December 2000. This was considered to be more than minor because of a delay in identifying and repairing sirens that would have been utilized to notify portions of the public in the event of a radiological emergency. However, there have been no significant problems with the sirens, the test results are in the green band for the siren testing performance indicator, and route alerting was available to compensate for any inoperable sirens. Under the significance determination process, the finding was considered to be of very low safety significance.
Inspection Report# : 2001007(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Preparedness Response Data System The team found that the Emergency Response Data System (ERDS) was found inoperable during an exercise in November 2000 and again during a test conducted in the 1st quarter 2001. The NRC conducted an ERDS test during this inspection and found both the system and its backup to be operable. This issue was determined to be of very low safety significance because the licensee retained capability to communicate via the telephone system. The failure to correct a deficiency identified during a drill/exercise is being treated as a non-cited violation of 10 CFR 50.47(b)(14).
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Operations Facility Inventory Records The licensee could not locate Emergency Operations Facility inventory records for the third quarter 2000 nor verify those inventories were actually conducted and a review of available quarterly inventory records identified cases where the records were not properly filled out. This issue was determined to be of very low safety significance because notwithstanding the discrepancies which were identified, the licensee had sufficient resources in the facilities to properly respond to an event. The failure to properly maintain emergency facilities and equipment is being treated as a non-cited violation of 10 CFR 50.47(b)(8) and the licensee's E-Plan, Section 8.3 which states quarterly inventories will be conducted.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct and/or document performance of quarterly communications links The licensee was not able to produce the 3rd quarter records for the operational check of the emergency communications links between facilities and could not verify that the tests had been conducted. This issue was determined to be of very low safety significance because the licensee had installed spare operable telephone lines. The failure to conduct and/or document the performance of quarterly communications tests is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.3 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 26 of 36 Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Emergency Response Organization Performance The team found that ten individuals assigned to the offsite and onsite monitoring teams had let their respirator qualifications lapse. This issue was determined to be of very low safety significance because there were sufficient responders with respiratory qualifications to fill the positions. The failure to maintain qualifications necessary to maintain proficiency as an emergency responder is being treated as a non-cited violation of 10 CFR 50.54(q) and Section 8.1.2 of the licensee's E-Plan.
Inspection Report# : 2001002(pdf)
Significance:        Feb 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish an effective emergency response training program The licensee continued to identify exercise deficiencies that are repetitive performance issues and are reflective of past performances, particularly in the area of plant assessment and the dissemination of the information to the general public. The team determined that the training program was not fully effective in preventing recurrence of repetitive exercise issues to ensure consistent emergency response organization performance. This issue was determined to be of very low safety significance because these performance issues did not deal with the risk significant planning standards (classifications, notifications, PARs). The failure to establish an effective training program to train employees and exercising, by periodic drills to ensure that employees maintain the proficiency of their specific emergency response duties, is being treated as a non-cited violation of 10 CFR Part 50.54(q) and Appendix E.IV.F.2.g.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to staff ENS line during event in a timely manner The licensee failed to establish a continuous communication line as requested by NRC. 10 CFR 50.72(c)(3) requires that during emergencies licensees maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC. The finding was treated as a non-cited violation of 50.72(c)(3) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Decrease in the effectiveness of the emergency plan The NRC identified a decrease in the effectiveness of the E-Plan because descriptions of some onsite ERO positions and the training program had been removed from the E-Plan. This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency plan content
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 27 of 36 The NRC identified that there was an inadequate description in the E-Plan of the joint news center (JNC) facilities and staff responsibilities and of the siren testing equipment used to verify siren operability. This finding was treated as a non-cited violation of 10 CFR 50 Appendix E requirements consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct ERO notification problems identified The NRC identified the failure to correct ERO notification deficiencies found as a result of drills or exercises as early as November 1999. Problems with the notification process still existed as demonstrated during the event of February 15, 2000, and as late as June 1, 2000, as evidenced by equipment reliability problems and inconsistent activation by assigned personnel. This finding was treated as a non-cited violation of 10 CFR 50.47(b)(14) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct off-hours exercise within six year period The licensee identified that they had not conducted an off-hours exercise at the required frequency. E-Plan Section 8.1.3, Drills and Exercises, commits the licensee to conduct an off-hours exercise once every six years. Prior to the February 15, 2000, event, the last off-hours exercise was conducted in 1993 and thus exceeded the six year periodicity.
This finding was treated as a non-cited violation of 10 CFR 50.54(q) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Significance:        Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to activate ERDS within one hour of an Alert During the February 15, 2000, event the licensee's failure to activate the Emergency Response Data System (ERDS) within one hour of an Alert was contrary to 10 CFR 50.72(a)(4). The ERDS was not made operable until approximately seven and one-half hours after the Alert declaration due to a problem with the telephone lines. This finding was treated as a non-cited violation of 10 CFR 50.72(a)(4) consistent with Section VI.A of the NRC Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance: N/A Nov 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.12.1 Violations of very low significance which were identified by the licensee have been reviewed by the inspector.
Corrective actions taken or planned by the licensee appear reasonable. These violations are listed in Section 4OA7 of this report
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 28 of 36 Inspection Report# : 2000013(pdf)
Public Radiation Safety Physical Protection Significance:        Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation UNIT 2 SECURITY RESPONSE FORCE MEMBER WAS FOUND INATTENTIVE TO ASSIGNED DUTIES On July 29, 2002, a member of the Unit 2 security response force was found inattentive to assigned duties. This inspector identified finding was treated as a non-cited violation of 10 CFR 73.55(b)(1)(i), and the Indian Point 2 Physical Security Plan. The security response force officer's inattentiveness to duties was determined to have very low safety significance, using the Interim Physical Significance Determination Process. The finding did not involve a significant compromise of the Physical Security Plan; no actual intrusion occurred; and, there have not been greater than two similar findings in the past four quarters.
Inspection Report# : 2002005(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control safeguards information The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-08 10CFR 73.21(a), Requirements for the protection of safeguards information requires, in part, "Each licensee....shall ensure that Safeguards Information is protected against unauthorized disclosure." In September, 2000, the improper handling of Safeguards documents was identified; as described in the licensee corrective action program, Reference Condition report 200007569.
Inspection Report# : 2000014(pdf)
Significance: N/A Jan 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to conduct adequate FFD testing The following finding of very low significance was identified by IP2 and is a violation of NRC requirements which meet Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCVs). NCV 05000247/2000-14-09 10CFR 26 Appendix A, Failure to Implement Requirements for FFD Testing. QA Annual Audit 00-04-D of the Fitness for Duty (FFD) Program identified that samples sent to the offsite lab for analysis were not tested to the correct criteria. Followup actions were appropriate. Reference Condition Report 200009066.
Inspection Report# : 2000014(pdf)
Miscellaneous Significance:        Aug 10, 2002
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 29 of 36 Identified By: NRC Item Type: NCV NonCited Violation FAIL TO USE THE APPROPRIATE TOOLING DEVICE FOR MOVEMENT OF FUEL ASSEMBLY G28 ON JULY 23, 2002 GREEN. On July 23, 2002, Entergy did not appropriately evaluate and implement short-term actions associated with Condition Report (CR) IP2-2002-07253. The consequence of the finding was the relocation of spent fuel assembly G-28 without the appropriate handling tools and precautions. The finding is more than minor since it could be reasonably viewed as a precursor to a significant event (dropped spent fuel assembly in the spent fuel pool). The Significance Determination Process is not modeled for a finding of this type. However, in accordance with NRC Manual Chapter 0612, this finding was reviewed by NRC risk analysts and management and has been determined to be of very low safety significance because no actual consequence existed and there was no unintended radiation worker exposure. The finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion V, and is being treated as a non-cited violation. (1R20)
Inspection Report# : 2002005(pdf)
Significance: N/A May 11, 2002 Identified By: Licensee Item Type: NCV NonCited Violation VIOLATION OF TECHNICAL SPECIFICATION 6.8.1.a - IMPROPER PROCEDURE USAGE On April 20, 2002, during a trip of one of the three condensate pumps, control room operators took incorrect action based on an abnormal operating instruction (AOI 21.1.1 step 5.6.4, by using a suction pressure number from this step that did not apply which resulted in their taking operator actions resulting in an unnecessary power transient. A May 8, 2002 condensate pump trip exemplified that this transient (a rapid down power) was not necessary to restore feedwater pump suction. The issue was more than minor since operator improper procedure usage is considered a precursor to a more significant event. Operator knowledge and skill performance issues have been captured in a number of individual NRC findings in past reports. Examples include operator re-qualification simulator test failures in September 2001 (reference NRC report 50-247/2001-013), and an overpower condition in August 2001 (reference NRC report 50-247/2001-09). The operator performance issues associated with the condensate pump trip were documented in the corrective action system as CRs 2000204180 and 200204183. Improper AOI 21.1.1 procedure usage was a violation of Technical Specification 6.8.1.a. This is being treated as a non-cited violation.
Inspection Report# : 2002003(pdf)
Significance: N/A Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES FOR SW VALVE LOCKING DEVICES A personnel error (human performance cross cutting issue) resulted in the failure to properly maintain locking devices on five service water test stop valves. The failure to maintain locking devices on service water valves per the operating procedures was a violation of Technical Specification 6.8.1.a. This is a non-cited violation.
Inspection Report# : 2002002(pdf)
Significance:      Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR REPEAT FAILURE OF SWN-7 The manual operator on service water (SW) valve SWN-7 failed on March 9 during operations to swap essential SW headers. SWN-7 is the isolation valve for the service water supply to turbine building loads. The inoperable valve could have resulted in insufficient service water flow for the emergency diesel generators and other safety systems had there been a demand for those safety systems. The operator on SWN-7 failed 6 other times since 1995. Following the early failures, an engineering evaluation determined that the design margin for the gear box in the manual operator was marginally adequate. Engineering work request 12110-99 was issued to replace the gear set on SWN-7 and similar valves with high strength materials. The engineering request was canceled in July 1999 and no action was taken. This issue had very low safety significance since the specific failure on March 9 and corrective actions occurred within the limiting condition for operation for the service water system, and no operating or stand-by mitigating equipment
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 30 of 36 supported by service water was called to perform its intended function. The failure to take adequate corrective action for repeat failures of service water valve SWN-7 was a violation of 10 CFR 50, Appendix B, Criterion XVI. This is being treated as a Non-cited violation.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation 10 CFR 50 APPENDIX B, CRITERION III, "DESIGN CONTROL" 10 CFR 50 Appendix B, Criterion III requires in part, that measures be established for the identification and control of design interfaces and for coordination among participating design organizations. The licensee did not ensure that the pressurizer level instrument drift evaluations were consistently bounded by the assumed instrument uncertainty within the safety analysis for a postulated Loss of Normal Feedwater event and a Loss of Offsite power event. The licensee documented this issue in condition report 2002000313.
Inspection Report# : 2002002(pdf)
Significance: N/A Feb 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH OVERPRESSURE PROTECTION SYSTEM The licensee's corrective actions in response to condition report 200004598 were untimely and ineffective to preclude the violation of TS figure 3.1.A-2. Condition report 200004598 initiated on June 16, 2000 identified that instrument uncertainty as stated in the TS basis was not incorporated in either the engineering analyses for the TS curves associated with heatup, cooldown and power operated relief valve setpoints, or the instrumentation for the power operated relief valve setpoints. The licensee failed to also consider the implication on the TS curves when overpressure protection system (OPS) is not considered operable and no reactor coolant system vent space exists. The corrective actions in response to this CR failed to preclude plant operations in violation of TS figure 3.1.A-2 on July 2, August 3, and November 30, 2000. This violation of 10 CFR 50 Appendix B, Criterion XVI had low actual safety significance because no consequence to the reactor coolant system pressure boundary occurred. This violation is being treated as a Non-cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001011(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: FIN Finding Failures during simulator exams - 2001 Licensee Operator Requalification Program The results of the 2001 Licensee Operator Requalification (LOR) Program showed a high number of crew and individual failures during the simulator exams. The licensee's preliminary investigation found the exam failures were caused by inadequate corrective actions and insufficient implementation of corrective actions for licensed operator knowledge and performance weaknesses identified during previous year LOR exams. The licensee determined the presently observed performance deficiencies were previously identified but not adequately corrected, aspects of which
 
3Q/2002 Inspection Findings - Indian Point 2                                                                Page 31 of 36 contributed to degraded performance in two plant reactivity management events and configuration control events in 2001. The inspector noted a root cause of the LOR program results (inadequate corrective actions) was also evident in recent plant events and NRC findings. This was an example of a cross cutting issue regarding human performance and problem resolution. Inspection Report 50-247/01-13 provides additional details regarding licensed operator requalification weaknesses.
Inspection Report# : 2001010(pdf)
Significance: N/A Dec 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Issue Condition Report and Implement Corrective Action as Required by 10 CFR 50, Appendix B, Criterion XVI The licensee's corrective actions in response to several equipment problems were ineffective. Repetitive failures of safety injection (SI) system relief valve, SI-855, and the low pressure steam dump valves were not prevented.
Appropriate analyses were not performed to fully understand the causes for the past failures. In addition, items related to these equipment problems were not entered in the corrective action program for resolution. This is a recurrent example of deficiencies in problem identification and resolution. The failure to correct conditions adverse to quality is considered a Severity Level IV violation of 10 CFR 50, Appendix B, Criterion XVI. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2001010(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: FIN Finding Identification of an Error in the Reactor Coolant System Activity Performance Indicator Data The inspector identified an error in the reactor coolant system (RCS) activity performance indicator (PI) data reported for the second quarter of 2001. Transcription errors and ineffective review contributed to the errant PI data. The errors had minimal significance since the PI remained within the green band. However, previous inspection findings identified errors in reporting Indian Point 2 PI data (reference NRC Inspections 05000247/00-01 and 00-11). This issue has more than minor significance because the failure to accurately report PI data potentially could impact the ability of the NRC to perform its regulatory function. The licensee entered this issue in the corrective action program as Condition Report 200109517.
Inspection Report# : 2001009(pdf)
Significance: N/A Oct 05, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Contrary to Criterion XVI The licensee corrective actions in response to past reactivity management and plant events were ineffective in precluding recurrent problems in log keeping, procedural adherence, and post-evolution debriefs. These deficiencies contributed to the August 17, 2001 overpower condition and the subsequent, untimely management review. This is a recurrent example of an issue in problem identification and resolution. The failure to correct conditions adverse to quality is considered a violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001009(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: FIN Finding Poor communications resulted in the untimely recognition of a degraded main turbine trip function Poor communications between plant operations staff and off-site electrical distribution personnel resulted in the untimely recognition of a degraded main turbine trip function that provided redundant protection from a fault in the offsite 345 kV system. Specifically, circuit troubleshooting in July 2001 identified a 345 kV pilot wire protection trip that was degraded since January 3, 2001. The licensee also identified poor quality drawings for offsite protection equipment and poor configuration control (a spare 125 volt DC breaker was open instead of closed as required).
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 32 of 36 Although the drawings and configuration control were not maintained by Indian Point Unit 2 personnel, they did impact the function of the electrical system as described in the UFSAR section 8.1.1 and 14.1.6.2. This issue was evaluated in the Significance Determination Process and found to have very low safety significance.
Inspection Report# : 2001008(pdf)
Significance: N/A Aug 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Condition Report pursuant to 10 CFR 50 Appendix B, Criterion XVI The licensee did not identify a condition adverse to quality evident in the repeated failures of a post-maintenance test (PMT) associated with the 23 emergency diesel generator (EDG). Following governor oil replacement in July 2001, the PMT was to perform the monthly surveillance PT-M21C, "Emergency Diesel Generator 23 Load Test." The procedure requires the EDG to be loaded to the 30 minute rating of 2300 kilowatts (kW). During the PMT, the 23 EDG could not achieve 2,300 kW, but was loaded to 2250 kW on July 25 and 2275 kW on July 26, 2001. The inability to reach desired loading was related to reaching terminal voltage limits when the EDG was tested with the generator operated in parallel with the offsite electrical grid. The licensee concluded that the inability to reach the desired load was an artifact of the test methodology and that the EDG would be able to reach the desired load under isochronous (loss of offsite power) conditions. Thus, the operability determination demonstrated the EDG could reach full load. Although EDG operability questions were addressed by this operability determination, the inspector was concerned with lack of progress in addressing this issue on previous occasions since six condition reports in the last three years documented EDGs not obtaining the desired loading due to offsite grid conditions (CR 199810268, 200003415, 200003494, 200003541, 200004426, 200004462). Previous corrective actions were not effective at resolving this testing deficiency. The failure to initiate a condition report for a condition adverse to quality (failure of a PMT for the EDG) is considered a violation of 10 CFR 50 Appendix B, criterion XVI. This violation is being treated as a Non-Cited violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
Inspection Report# : 2001008(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding ASSESSMENT OF WORK ON THE STATION AUXILIARY TRANSFORMER (SAT) TAP CHANGER Con Edison's assessment of the work on the station auxiliary transformer (SAT) tap changer indicated the maintenance had high risk significance due to the potential for a plant transient and electrical system perturbations. Weaknesses were noted in the initial work planning when the tap changer maintenance was attempted on June 7. During the pre-job brief, control room operators identified problems in implementing contingency actions and requested additional contingency planning. Con Edison subsequently refined the risk assessment, implemented planning details, and completed the tap changer maintenance on the on June 19, 2001 with a daily risk factor comparable to the baseline value. The failure to initially manage plant risk during the maintenance activity was a contributor to an adverse trend in problem identification and resolution.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO ADEQUATELY CONTROL TAGGING ACTIVITIES While Gas Turbine GT1 was out of service for repairs, Con Edison applied a tagging order to de-energize electrical equipment prior to asbestos abatement. The tagging order caused the inadvertent loss of IP1 DC control power which impacted the ability to electrically operate 13.8 KV breakers that supply alternate safe shutdown power to IP2 safety systems. The over current protection intended to protect the safe shutdown equipment from a fault was unavailable for about 6 hours. The adequacy of IP1 electrical drawings and staff knowledge of available drawing resources were a factor in the tagging problem. Con Edison identified other inadequacies in IP1 electrical drawings and equipment labeling during the period which impacted tagging activities. The failure to adequately control tagging activities was a contributor to an adverse performance trend in human performance.
Inspection Report# : 2001006(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 33 of 36 Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: FIN Finding SEVERAL EVENTS THAT WERE INDICATIVE OF AN ADVERSE TREND IN HUMAN PERFORMANCE Several other events during the period were indicative of an adverse trend in human performance, including operator performance following the June 5 fire system leak into the utility tunnel; the conduct of a reactor protection system test with an unqualified technician; inadequate preparation resulting in an unnecessary 100 mRem radiation exposure; and, work on the wrong emergency battery light. In response, Con Edison reset the "event free clock" and conducted a station stand down on June 14 - 15, 2001 to review human performance issues.
Inspection Report# : 2001006(pdf)
Significance: N/A Jul 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE PROTECTION DESIGN BASIS COMBUSTIBLE LOADING The inspector identified during a review of the fire hazards analysis that each fire zone throughout the plant did not have a retrievable basis for their combustible loading. The failure to provide a design basis for combustible loading was contrary to TS 6.8.1.a and License Condition 2.K. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25368).
Inspection Report# : 2001006(pdf)
Significance:      Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation During implementation of a plant modification, workers failed to perform a work area walkdown, pre-job brief, and review of removal drawings The licensee issued a modification to reroute the nitrogen piping to the reactor coolant drain tank. During implementation of the modification, workers failed to review drawings, perform a work area walkdown, and conduct a pre-job brief. The workers failed to locate the correct pipe and cut the nitrogen supply line to the safety injection accumulators and the power operated relief valves. This issue had very low safety significance because the safety injection accumulators and the power operated relief valves were not required to be operable at the time. The failure to implement maintenance procedures pursuant to technical specification 6.8.1 is being treated as a non-cited violation.
Inspection Report# : 2000011(pdf)
Significance:      Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation A minor fire inside containment occurred due to a failure to properly evaluate and control transient combustibles during a grinding evolution A minor fire inside containment occurred on September 3, 2000, when sparks from a grinding evolution landed on a combustible foreign material exclusion (FME) tarp during work controlled under work permit 1060, "Install Reactor Cavity Decking." The fire occurred due to the failure to properly evaluate and control transient combustibles. This issue had very low safety significance because the location of the fire did not impact safe shutdown equipment. The failure to control transient combustibles in accordance with station administrative orders is being treated as a non-cited violation of license condition 2.K.
Inspection Report# : 2000011(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: FIN Finding Operations and Engineering support areas, corrective actions to resolve known problems were untimely and incomplete.
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 34 of 36 In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability.
Inspection Report# : 2000007(pdf)
Significance:      May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly initiate CRs and initiate appropriate corrective actions Con Edison did not properly disposition or enter some conditions adverse to quality into their corrective action program as required by procedure. A selected review of the Communications to Staff (CTS) database, a database of procedure enhancement recommendations, determined that one CTS item was not adequately resolved and two additional CTS items met the threshold for initiating a condition report (CR) for which a CR was not initiated. This non-cited violation is associated with the failure to initiate condition reports as required by Con Edison's procedures. The issue was determined to be of very low risk significance, because the most notable problem was related to a delay in reducing plant pressure, and did not result in any appreciable increase in the release of activity during the steam generator tube failure event.
Inspection Report# : 2000007(pdf)
Significance: N/A May 26, 2000 Identified By: NRC Item Type: NCV NonCited Violation Faulure to follow procedures and enter the required data into the control room log The control room operators did not enter significcant plant items, such as event declaration and implementaiton of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999 loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones, it was considered important because prior corrective actions were not effective.
Inspection Report# : 2000007(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Container Liner Degradation The containment liner became corroded due to prolonged contact with borated water in areas where moisture barriers were degraded. Con Edison actions continued to investigate and repair liner degradation, and to assure that margins to design limits were maintained.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 35 of 36 Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Conclusions for Maintenance and Surveillance Maintenance activities were satisfactorily completed. The conduct of surveillance tests during the period was acceptable. Maintenance and test activities were not consistently performed in accordance with expectations and administrative controls. The initial evaluations in preparation for a turbine load test did not completely consider shutdown risk.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Nuclear Facilities Safety Committee Plant management presentations to the Nuclear Facilities Safety Committee were incomplete. However, the committee members appeared well prepared and provided good discussions on the February 15 steam generator tube leak event.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Review of Response to Loss of Power and Air Supply to Steam Generator Nozzle Dams The operators promptly responded to the loss of power to the steam generator nozzle dams. The nozzle dam normal air supply was lost; however, no loss of reactor coolant system inventory occurred, and no monitoring existed for the nozzle dams for approximately one hour. Con Edison failed to control and integrate several temporary facility changes for the nozzle dam support systems. Inadequate coordination between operators and workers resulted in a near miss for a significant injury Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Self Disclosing Item Type: FIN Finding Steam Generator Examinations Steam generator eddy current testing and analysis was conducted. The eddy current test results revealed defects which resulted in a Classification of C-3 per Technical Specifrication 4.13. More detailed review of steam generator inspecction results is under the purview of the NRC Office of Nuclear Reactor Regulation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES ON CORE DIFFERENTIAL TEMPERATURE The operators failed to control RCS differential temperature within limits during RHR system operation. The failure to follow SOP 4.2.1 was a non-cited violation of NRC requirements. Licensee actions continued at the end of the
 
3Q/2002 Inspection Findings - Indian Point 2                                                                  Page 36 of 36 inspection period to evaluate the impact on the baffle-former and baffle-barrel bolts in the reactor vessel internals, and to resolve this matter prior to plant restart.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET APPENDIX R FOR OIL COLLECTION SYSTEM The failure to collect leakage from the vent pipe and the lower oil reservoir drain connections on three RCP motors is considered a violation of 10 CFR 50 Appendix R, Section III. This Severity Level IV violation is being treated as a Non-Cited Violation. A long-standing deficiency in the oil collection system went uncorrected.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: AV Apparent Violation FAILURE TO MEET IVSWS LICENSING BASIS Con Edison did not recognize a long-standing difference between the design and licensing basis for the isolation valve seal water system. Despite several past events and a design basis verification program which highlighted IVSWS performance issues, Con Edison failed to correct a basic design deficiency and assure that the licensing basis was met.
Operability evaluations were less than adequate and corrective actions were narrow and untimely. The failure to assure regulatory requirements were correctly translated into specifications, drawings and procedures was an apparent violation.
Inspection Report# : 2000003(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE INSTRUCTIONS FOR FIRE DAMPERS lack of maintenance installation instructions contributed to the failure of cable spreading room fire dampers to fully close. The faulty dampers caused the suppression system to be degraded for approximately 3 months. The failure to maintain provisions of the NRC-approved fire protection plan as described in the UFSAR and approved NRC Safety Evaluation Report is a Non-Cited Violation.
Inspection Report# : 2000003(pdf)
Last modified : December 02, 2002
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                            Page 1 of 9 Indian Point 2 Initiating Events Significance:        Aug 10, 2002 Identified By: Self Disclosing Item Type: FIN Finding CONTRACTOR WORKED OUTSIDE HIS ESTABLISHED JOB SCOPE FOR LANDSCAPING ACTIVITIES On July 19, 2002, a contractor worked outside his established job scope for landscaping activities. The consequences of this human performance error were the accidental electrocution of the individual and an offsite power electrical transient (loss of the 138 kilovolt station auxiliary transformer for approximately seven hours). This partial loss of offsite power event was more than minor, in that it impacted the reactor safety cornerstone with respect to the initiating event objective of limiting the likelihood of an event that upsets plant stability and challenges the critical safety function of the on-site emergency diesel generators. Notwithstanding the loss of life (which the Department of Labor, Occupational Safety and Health Administration is reviewing), this electrical transient event was of very low safety significance because it did not contribute to the likelihood of: loss of coolant accidents, a reactor trip and the unavailability of accident mitigation equipment or functions being unavailable; or of a fire or internal/external flood. No violations of NRC requirements were identified.
Inspection Report# : 2002005(pdf)
Significance:        May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding REDUCTION OF PLANT POWER BY CONTROL ROOM OPERATORS DUE TO CONDENSATE PUMP MOTOR FAILURES On April 20, 2002, and on May 8, 2002, the control room operators reduced plant power due to condensate pump motor failures. A lack of a predictive maintenance program and an improperly set oil level indication system were the causes for two separate condensate motor failures.
The events are more than minor since both events increased the likelihood of an initiating event. Operator response was necessary to ensure an automatic reactor trip did not occur due to a low steam generator level. The performance issues were of very low safety significance since there was no impact to normally available mitigating equipment.
Inspection Report# : 2002003(pdf)
Significance:        Mar 30, 2002 Identified By: NRC Item Type: FIN Finding INAPPROPRIATE PROCEDURE FOR INOPERABLE STATION AUXILIARY TAP CHANGER The procedure in use was inappropriate in that it did not require that the 138 kilovolt off-site power system be declared inoperable during scheduled maintenance on the station auxiliary transformer (SAT) tap changer. On February 28, 2002, for approximately 51 minutes, control room operators had placed the SAT tap changer in manual and local control in accordance with system operating procedure (SOP) 27.1.7, "Operation of Main, Station and Unit Auxiliary Transformers," section 4.8. The scheduled maintenance was not intrusive into tap changer operation, however, the licensee had not fully evaluated if the intended function could be maintained with operator compensatory actions to restore the tap changer to automatic. The limiting condition for operation in technical specification 3.7.B.3 for a loss of the 138 kilovolt power system is 24 hours, which was not exceeded during this scheduled maintenance activity. The issue had a credible impact on safety.
Inappropriate control of the SAT tap changer impacts the initiating event cornerstone in that a loss of off-site power is more likely following a reactor trip. This issue was determined to be of very low safety significance (Green) using phase one of the SDP because no reactor trip occurred during the inspection period and no mitigating systems were directly impacted by the maintenance on the SAT tap changer.
Inspection Report# : 2002002(pdf)
Mitigating Systems Significance:        Dec 28, 2002 Identified By: NRC Item Type: FIN Finding UNTIMELY OPERABILITY DETERMINATION FOR THE 21 DIESEL GENERATOR
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                          Page 2 of 9 On October 9, 2002, the licensee's organization did not identify in a timely manner that the 21 emergency diesel generator was inoperable. The causes for the untimely operability evaluation were fragmented communications between Entergy departments, untimely drip tank sample results, system engineering turnover, and a lack of sensitivity to a loss of the emergency power source safety function. The time between when the non-licensed operator had reported and added inventory to the jacket water expansion tank to the time the emergency diesel generator was declared inoperable was 7.5 hours which exceeded the limiting condition for operation within TS 3.0.1 to be in hot shutdown within seven hours. In the absence of reasonable expectation that a component is operable, the component shall be declared inoperable immediately. The untimely operability evaluation affects the mitigating systems cornerstone objective. The attribute is human performance pre-event. This finding is of very low safety significance in phase 1 of the SDP since the 21 EDG was subsequently declared inoperable and actions within the TS were adhered to. This finding did not result in an actual loss of the emergency on-site power source safety function nor did it increase the risk significance for external events. No violations of NRC requirements were identified.
Inspection Report# : 2002007(pdf)
Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE CAUSE OF 23 EDG OUTPUT BREAKER TO CLOSE The inspector identified that Entergy did not adequately evaluate the cause blown control power fuses on the 23 EDG output breaker cubicle on November 10, 2002, that subsequently caused the breaker from closure on November 14, 2002. On November 14, 2002, Entergy identified the cause of the breaker failure as improper operation of the inertial latch mechanism. This performance issue is being treated as a Non-cited Violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is more than minor because the failure to identify the cause and preclude recurrence was considered a precursor to a more significant safety issue, in that, the plant could have started up with only 2 available EDGs - a violation of TS - and not have know it for approximately one month. The issue was determined to be of very low safety significance (Green) in accordance with MC 0609 Appendix G, since greater than three offsite and onsite power sources were available to cope with a postulated loss of offsite power.
Inspection Report# : 2002007(pdf)
Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE POST WORK TEST ON STEAM STOP CHECK VALVE The post work test on the 22 steam generator stop check valve (MS-41) failed to identify that the valve plug was installed upside down. This self-revealing event was identified on November 20, 2002, when operators responded to steam leak-by from this tagged closed valve that resulted in a fire alarm in the auxiliary feedwater pump room. This finding is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," in that, the post work test, PT-R67A, Reverse Flow Check at MS-41 and MS-42 Alternate Test, revision 1 did not adequately verify that MS-41 was properly reinstalled after preventative maintenance. The performance finding is considered more than minor, since the improperly installed valve plug would not have been identified prior to auxiliary feedwater system operability, had it not been identified during an unrelated tagout on the steam supply to the 22 auxiliary feedwater turbine. This is considered very low risk significance in accordance with NRC MC 0609 Appendix G since two alternate core cooling paths were available.
This is an example of insufficient Entergy oversight of contractor work activities.
Inspection Report# : 2002007(pdf)
Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONFIGURATION CONTROL FOR A SAFETY RELATED SYSTEM The inspector identified an example of inadequate configuration control for a safety-related system. On November 20, 2002, the inspector identified that two 125 vdc circuit breakers were in their correct position (open) but administrative locking devices were not installed. The breakers are used to cross-connect the 21 and 22 125 vdc buses. This is considered a Non-Cited Violation of Technical Specification 6.8.1.a.,
which includes requirements for procedure adherence and operations of safety-related systems including the 125 volt DC system. Check off list (COL) 27.1.6, Instrument Buses, DC Distribution and PA Inverter, revision 18 requires the breakers to be open and locked. This performance deficiency is more than minor since more than one breaker was in the required position, but not locked. The finding impacts the mitigating systems cornerstone and is associated with pre-event human error. The finding is considered very low safety significance (Green) since the operability or availability of the 21 and 22 DC buses were not impacted.
Inspection Report# : 2002007(pdf)
Significance:        Sep 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                            Page 3 of 9 NON-CITED VIOLATION OF TS 6.8 INVOLVING DEFICIENT GUIDANCE IN PROCEDURE AOI 27.1.1 Abnormal Operating Instruction (AOI) 27.1.1, "Loss of Normal Station Power," was deficient, in that no steps were provided in the procedure to identify that the lockout relays for the component cooling water (CCW) pumps were required to be reset following a loss and restoration of power to the motor supply breakers. This deficient procedure is being treated as a Non-Cited Violation of Technical Specification (TS) 6.8, "Procedures and Programs," in accordance with the NRC Enforcement Policy. The consequence of this finding was that the pump lockout relays would have prevented the 21 and 23 CCW pumps from starting automatically on low CCW system header pressure, for 12 days and 21 days, respectively. This finding represented a partial loss of the CCW system function and would reasonably have been corrected by operator action.
Inspection Report# : 2002006(pdf)
Significance:        Aug 10, 2002 Identified By: NRC Item Type: FIN Finding OPERATORS DID NOT IDENTIFY THE APPLICABILITY OF A SHUTDOWN TECHNICAL SPECIFICATION On July 19, 2002, operators did not identify the applicability of a shutdown Technical Specification (TS) associated with the planned removal from service of the 22 emergency diesel generator (EDG) while the 138 kilovolt off site power system was still out-of-service. This finding was associated with the reactor safety cornerstone with respect to the mitigating systems objective of ensuring the availability, reliability and capability of the EDG to respond to initiating events, such as a loss of offsite power, to prevent undesirable consequences. No violation of NRC requirements was identified, since Entergy restored the 22 EDG prior to exceeding the allowed outage time per TS 3.0.1. This finding was of very low safety significance since it did not represent a total loss of emergency power safety function.
Inspection Report# : 2002005(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902. The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program. Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls. [Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation TURBINE DRIVEN AUX FEED PUMP OIL ISSUES The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning three issues with the control and monitoring of lubrication oil used on the turbine driven auxiliary boiler feed water pump (22 ABFP). Each issue involved incomplete evaluations that led to repeat problems and potential for pump damage. The evaluation and corrective actions following identification in February 2002 that the wrong oil was added to the turbine speed governor were not fully effective. The evaluation of this issue identified that operators were not logging the quantity or specification of oil added during rounds or operation of equipment, but no actions were taken to address the issue. Additionally, the team noted that on July 10, while preparing to run the pump, Entergy identified additional confusion regarding the specification of oil to be added to the governor, an issue that should have been resolved. Station personnel did not identify that oil analysis results in May 2002 showing a decrease in oil viscosity indicated that the wrong oil was likely added to a pump bearing and that corrective actions for a similar problem previously identified in May 2001 were ineffective. The evaluation and corrective actions following identification in October 2001 of issues with the required oil level in the pump inboard bearing were not fully effective, specifically the design drawing, the vendor manual, and operator training contained inconsistent information. These issues were evaluated using Phase I of the NRC SDP to have very low safety significance (Green), because pump operability was not directly affected. These issues are being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because the
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                            Page 4 of 9 issues have been entered into Entergy's CAP. (NCV 50-247/02-010-002 Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SETPOINT DATABASE NOT CORRECTED FOR CIRCUIT BREAKER OVERCURRENT PROTECTION DEVICE SETPOINTS The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to promptly identify, determine the cause, and correct circuit breaker amptector setpoint database errors. The control of design setpoints is necessary to ensure the availability, reliability and capability of safety-related electrical systems. This issue was evaluated using Phase I of the NRC SDP and determined to have very low safety significance (Green), because the team did not identify any instances where a circuit breaker would not have been able to perform its safety function. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP. (NCV 50-247/02-010-003)
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SAFETY INJECTION TOPPING PUMP VIBRATION CONSEQUENCES TO SAFETY-RELATED PIPING The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to identify that vibration of the non-safety-related SI accumulator topping pump caused stresses in adjacent safety-related piping that were above the code allowable values. The team evaluated this issue using Phase I of the NRC SDP, determining it to have very low safety significance (Green),
because liquid penetrant examinations in the areas of high stress did not identify any piping damage. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP. (NCV 50-247/02-010-004 Inspection Report# : 2002010(pdf)
Significance:        Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding MULTIPLE GROUNDS ON THE PROTECTIVE CIRCUIT FOR UNIT 1 SUBSTATION 102NS3 RESULTED IN A LOSS OF THE 13.8 KILOVOLT LIGHTING AND POWER BUS SECTION 3 On May 17, 2002, multiple grounds on the protective circuit for Unit 1 substation 102NS3 resulted in a loss of the 13.8 kilovolt (kv) lighting and power bus section 3. The consequence of this event was a loss of alternate safe shutdown power to all major alternate safe shutdown pumps and selected instrumentation. At the time, the Unit 2 normal and emergency electrical power supplies were available to supply power to the above stated mitigation equipment and instrumentation. The licensee repaired and restored the 13.8 kv bus section 3 within 30 hours of the fault. The performance issue is inadequate retirement of protective circuits for 440 volt substations (132PC3 and 142PC3) that could impact availability of alternate safe shutdown power supplies. This issue is more than minor since unavailability of alternate safe shutdown equipment for 30 hours is viewed as a precursor to a significant event and the alternate safe shutdown power supplies are a risk-significant maintenance rule system which was unavailable for greater than 24 hours.
Inspection Report# : 2002004(pdf)
Significance:        Feb 09, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR A TEMPORARY FACILITY CHANGE INVOLVING THE AUXILIARY FEEDWATER SYSTEM BACKUP NITROGEN SUPPLY SYSTEM.
The inspector identified that a temporary facility change (TFC) for the backup auxiliary feedwater system (AFW) nitrogen supply was deficient because component specifications critical to the design were not identified in the design package. This issue was considered more than minor because of the potential for an improper component substitution to impact operability of a risk significant system. However, this issue was determined to be of very low safety significance using phase one of the SDP because the modification was adequate as installed. The failure to include design specifications in the TFC was a violation of Criterion III, Design Control. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance:        Nov 05, 2001 Identified By: NRC
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                          Page 5 of 9 Item Type: FIN Finding CREW HIGH FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with crew high failure rate (four of seven crews failed) during facility-administered annual licensed operator requalification examinations conducted last fall. The finding was previously characterized as having substantial safety significance (Yellow) in NRC Inspection Report 50-247/01-13. The inspectors noted that the licensee's evaluation identified a fundamental underlying weakness: The station has yet to overcome cultural weaknesses that include an unwillingness to confront poor performance, an over reliance on procedures to change behavior, and compartmentalization. More specifically, the licensee identified three root causes: 1) Operations training had not focused on the basic building blocks that ensure a healthy program; 2)
The station had not maintained a core of career oriented, plant knowledgeable instructors and operators; and 3) Operations department involvement with Operations Training had often been ineffective. The inspectors concluded that the methodology and level of detail of the licensee's root cause evaluation were reasonable. The licensee implemented a number of corrective actions to address the identified causes. The corrective actions are described in the station's Training Improvement Plan. The more significant corrective actions included initiatives that aimed to 1) improve the quality of training and training materials; 2) increase the number of instructors who have Unit 2 plant experience; and
: 3) provide additional management support and oversight of training. The inspectors determined that the corrective actions are appropriately focused on the identified causes. These actions were appropriately prioritized, and either complete or scheduled for completion. Notably, the licensee took strong immediate corrective actions following the requalification examination failures to provide extensive retraining to each shift, and continue to provide this high intensity training. The inspectors independently assessed the extent of the underlying conditions that led to the Yellow finding and found that performance issues had also existed in other Operations Training programs, such as initial licensed operator and non-licensed operator training programs. These problems existed for at least three years, both prior to and following the steam generator tube failure event in 2001. Although licensee audits and assessments had identified most of the performance problems prior to the crew high failure rate, they did not identify long-term operator performance as a concern. The inspectors concluded that the licensee's extent of condition review appropriately bounded the underlying conditions that led to the Yellow finding as evidenced by the fact that the licensee had also identified the duration and extent of the problems, and the failure to recognize the long standing issues. (Updated) FIN 05000247/01-013-01: Proposed finding due to crew high failure rate during the 2001 annual requalification simulator examinations. This finding was documented in an October 2001 inspection and initially characterized as a potential Yellow finding, the final safety significance to be determined (TBD).
This finding was subsequently evaluated under the significance determination process (SDP) and characterized as (reference NRC to Entergy letters dated December 5, 2001, and February 28, 2002). The 95002 Supplemental Inspection (reference Inspection Report No. 50-247/02-09, dated May 31, 2002), assessed the licensee's evaluation of the crew high failure rates and the corrective actions taken to address this performance issue. As stated in the cover letter to Inspection Report No. 50-247/02-09, this finding remains open until after the completion of Entergy's licensed operator requalification examinations, scheduled for September-October 2002, and further review by the NRC. This item remains open.
Inspection Report# : 2001013(pdf)
Inspection Report# : 2002004(pdf)
Inspection Report# : 2002009(pdf)
Significance:        Apr 13, 2001 Identified By: NRC Item Type: URI Unresolved item Adequacy of Hemyc Cable Wrap Fire Barrier Qualification Test and Evaluation Based on the review of test reports CTP-1026 and CTP-1077, the team determined that the results of the engineering test alone were inconclusive for qualifying the fire barrier system as a one hour rated fire barrier. The team noted that ConEd had only credited the Hemyc fire barrier on the 23 ABFP for 30 minutes, however, due to identified test discrepancies, the 30 minute rating was also inconclusive. This issue is unresolved pending further NRC review to determine whether the qualification tests of the Hemyc fire barrier wrap systems are acceptable.
Inspection Report# : 2000004(pdf)
Barrier Integrity Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation OPERATORS DEVIATE FROM PLANT OPERATING PROCEDURES On November 23, 2002, during a plant cooldown, Entergy deviated from the guidance of plant operating procedure (POP) 3.3, Plant Cooldown, Rev. 57. The consequence of the failure to follow the POP guidance was to exceed the operational limits on the steam generator tube sheet differential pressure of 1600 psid with a maximum value of approximately 1855 psid. Control room operators were unaware of this operational limit. Reviews of steam generator manufacturer specifications and the Updated Final Safety Analysis Report design basis accident analysis information indicated that the steam generator tubes were designed to withstand up to 2485 psid during upset and hydrostatic conditions. Therefore, the structural integrity and qualification of the steam generator tubes was maintained. Failure to document the basis of marking non-conditional steps in POP 3.3 as not-applicable is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V. This issue was considered more than minor because it represented a lack of understanding of procedure requirements and awareness of plant
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                        Page 6 of 9 operational limitations. This finding is considered very low safety significance (Green) in accordance with manual chapter 0609 Appendix G, in that the core cooling pathway via the steam generators was not impacted.
Inspection Report# : 2002007(pdf)
Significance:        Jun 29, 2002 Identified By: NRC Item Type: FIN Finding DURING SURVEILLANCE TESTING OF THE SAFETY INJECTION DISCHARGE MOTOR-OPERATED VALVE 851B, THE VALVE FAILED TO STROKE CLOSED On May 27, 2002, during surveillance testing of the safety injection discharge motor-operated valve (851B), the valve failed to stroke closed.
The initial operability evaluation did not consider the non-automatic containment isolation function for this valve. This event was documented in condition report No. 200205433. The performance issue associated with this finding is a weakness in operator knowledge of multi-function safety system components. This is the second recent example where operators did not consider this function for a safety-related valve. The first example was documented in NRC report 50-247/2002-003, section 1R15. The untimely and incomplete operability assessment for safety injection discharge valve 851B has very low safety significance since the containment isolation valve was restored to an operable status prior to exceeding Technical Specification 3.6.A.3.a.2.d limiting condition for operation.
Inspection Report# : 2002004(pdf)
Significance:        May 11, 2002 Identified By: Licensee Item Type: FIN Finding UNTIMELY OPERATOR EVALUATION FOR CONTAINMENT ISOLATION VALVE 869B On April 11, 2002, operators did not complete a timely operability evaluation for containment isolation valve 869B after the disconnect switch operating handle on motor control center (MCC)26BB broke while applying an equipment tagout. At the time, the operators neither verified that the disconnect would operate nor completed an adequate evaluation regarding the ability to close valve 869B to perform its containment isolation function. An operability evaluation was completed about six hours later by a different operating crew and the operators then entered a four-hour limiting condition for operation and isolated the containment penetration per the technical specifications 3.6.A.3.a.2.b. The untimely operability evaluation increased the unavailability time for the containment spray system. The inoperable containment isolation valve issue was more than minor because it impacts the containment barrier. This issue had very low safety significance since the containment isolation valve was repaired and restored to an operable status prior to exceeding technical specification 3.6.A.3.a.2.d. This issue was an example of untimely operator implementation of technical specification requirements in response to degraded safety equipment.
Inspection Report# : 2002003(pdf)
Significance:        Feb 09, 2002 Identified By: Licensee Item Type: NCV NonCited Violation POSTULATED CONTAINMENT LEAKAGE IN EXCESS OF TS 3.6 LIMITS The licensee identified a minor leak in the service water piping while the plant was in cold shutdown for a maintenance outage. The leak was repaired prior to startup, and an extent of condition review identified no other defects in service water piping. The licensee determined that the leak most probably initiated during the shutdown period; however, for significance determination the licensee postulated that the defect existed during plant operation prior to the outage in order to conservatively estimate containment leakage during design basis events. This issue was determined to be more that minor because the defect in the service water piping created a potential leakage path from containment. However, the issue was considered to be of very low safety significance using phase two of the SDP because the service water leak did not affect the function of safety equipment, and the containment leakage potential was significantly less than that which would result in a large early release.
The failure to maintain containment integrity was a violation of TS 3.6. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Emergency Preparedness Significance:        Sep 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECT PREVIOUSLY IDENTIFIED CONDITION IN THE JNC REGARDING THE TIMELY AND ACCURATE DISSEMINATION OF INFORMATION During the emergency plan exercise conducted on September 24, 2002, the licensee JNC personnel proceeded with a 1:55 p.m. press
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                        Page 7 of 9 conference informing the media that no release was in progress when, according to the exercise events, a release had begun just prior to the press conference. This information was not in conflict with the emergency alert system message that was in effect at the time of the briefing.
However, the failure to provide updated information could cause confusion for those receiving it through the media. This failure was a previously identified weakness as documented in Drill Critique Reports and in condition reports. This is a non-cited violation of 10 CFR 50 Appendix E Section IV.F.2.g which requires, in part, that weaknesses or deficiencies that are identified during a drill or exercise shall be corrected. The risk associated with this release of incorrect information was determined to be of very low significance because it does not constitute a loss of function in meeting the applicable planning standard (10 CFR 50.47(b)(14).
Inspection Report# : 2002012(pdf)
Significance: N/A May 11, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF 10 CFR 50.54(q) FOR ACCOUNTABILITY On March 6, 2002, the licensee implemented changes to the accountability process that decreased the effectiveness of the Emergency Plan (E-Plan). The finding was considered more than minor because, if left uncorrected, it would become a more significant safety concern. Changing commitments in the E-Plan without prior approval impacts the NRC's ability to perform its regulatory function and potentially creates an ineffective response to a radiological emergency. The consequences of this change were minimal because it did not preclude the function of accountability from being performed, albeit delayed. The licensee has implemented the corrective actions and has since met the timeliness goal.
This change which decreased the effectiveness of the Plan is being treateed as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued May 1, 2000.
Inspection Report# : 2002003(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Significance:        Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation UNIT 2 SECURITY RESPONSE FORCE MEMBER WAS FOUND INATTENTIVE TO ASSIGNED DUTIES On July 29, 2002, a member of the Unit 2 security response force was found inattentive to assigned duties. This inspector identified finding was treated as a non-cited violation of 10 CFR 73.55(b)(1)(i), and the Indian Point 2 Physical Security Plan. The security response force officer's inattentiveness to duties was determined to have very low safety significance, using the Interim Physical Significance Determination Process. The finding did not involve a significant compromise of the Physical Security Plan; no actual intrusion occurred; and, there have not been greater than two similar findings in the past four quarters.
Inspection Report# : 2002005(pdf)
Miscellaneous Significance:        Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAIL TO USE THE APPROPRIATE TOOLING DEVICE FOR MOVEMENT OF FUEL ASSEMBLY G28 ON JULY 23, 2002 GREEN. On July 23, 2002, Entergy did not appropriately evaluate and implement short-term actions associated with Condition Report (CR)
IP2-2002-07253. The consequence of the finding was the relocation of spent fuel assembly G-28 without the appropriate handling tools and precautions. The finding is more than minor since it could be reasonably viewed as a precursor to a significant event (dropped spent fuel assembly in the spent fuel pool). The Significance Determination Process is not modeled for a finding of this type. However, in accordance
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                          Page 8 of 9 with NRC Manual Chapter 0612, this finding was reviewed by NRC risk analysts and management and has been determined to be of very low safety significance because no actual consequence existed and there was no unintended radiation worker exposure. The finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion V, and is being treated as a non-cited violation. (1R20)
Inspection Report# : 2002005(pdf)
Significance: N/A Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Improved performance in the areas of design control, equipment and human performance, and corrective actions.
Overall, the team found that Entergy operated IP2 safely and that through implementation of the FIP, progress had been made in improving performance in the areas of design control, equipment and human performance, and corrective actions, Specifically, the team determined overall: quality of engineering products has improved, and design and licensing basis control have been strengthened; equipment performance improved (including reduced backlogs of corrective maintenance work orders, operator workarounds, CCR deficiencies, etc); station human error rate and number of equipment mis-positioning events have declined; and, improved effectiveness of corrective action program, including identification and documentation of issues at a low threshold.
Inspection Report# : 2002010(pdf)
Significance: N/A May 11, 2002 Identified By: Licensee Item Type: NCV NonCited Violation VIOLATION OF TECHNICAL SPECIFICATION 6.8.1.a - IMPROPER PROCEDURE USAGE On April 20, 2002, during a trip of one of the three condensate pumps, control room operators took incorrect action based on an abnormal operating instruction (AOI 21.1.1 step 5.6.4, by using a suction pressure number from this step that did not apply which resulted in their taking operator actions resulting in an unnecessary power transient. A May 8, 2002 condensate pump trip exemplified that this transient (a rapid down power) was not necessary to restore feedwater pump suction. The issue was more than minor since operator improper procedure usage is considered a precursor to a more significant event. Operator knowledge and skill performance issues have been captured in a number of individual NRC findings in past reports. Examples include operator re-qualification simulator test failures in September 2001 (reference NRC report 50-247/2001-013), and an overpower condition in August 2001 (reference NRC report 50-247/2001-09). The operator performance issues associated with the condensate pump trip were documented in the corrective action system as CRs 2000204180 and 200204183.
Improper AOI 21.1.1 procedure usage was a violation of Technical Specification 6.8.1.a. This is being treated as a non-cited violation.
Inspection Report# : 2002003(pdf)
Significance: N/A Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES FOR SW VALVE LOCKING DEVICES A personnel error (human performance cross cutting issue) resulted in the failure to properly maintain locking devices on five service water test stop valves. The failure to maintain locking devices on service water valves per the operating procedures was a violation of Technical Specification 6.8.1.a. This is a non-cited violation.
Inspection Report# : 2002002(pdf)
Significance:        Mar 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR REPEAT FAILURE OF SWN-7 The manual operator on service water (SW) valve SWN-7 failed on March 9 during operations to swap essential SW headers. SWN-7 is the isolation valve for the service water supply to turbine building loads. The inoperable valve could have resulted in insufficient service water flow for the emergency diesel generators and other safety systems had there been a demand for those safety systems. The operator on SWN-7 failed 6 other times since 1995. Following the early failures, an engineering evaluation determined that the design margin for the gear box in the manual operator was marginally adequate. Engineering work request 12110-99 was issued to replace the gear set on SWN-7 and similar valves with high strength materials. The engineering request was canceled in July 1999 and no action was taken. This issue had very low safety significance since the specific failure on March 9 and corrective actions occurred within the limiting condition for operation for the service water system, and no operating or stand-by mitigating equipment supported by service water was called to perform its intended function. The failure to take adequate corrective action for repeat failures of service water valve SWN-7 was a violation of 10 CFR 50, Appendix B, Criterion XVI. This is being treated as a Non-cited violation.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation 10 CFR 50 APPENDIX B, CRITERION III, "DESIGN CONTROL" 10 CFR 50 Appendix B, Criterion III requires in part, that measures be established for the identification and control of design interfaces and for coordination among participating design organizations. The licensee did not ensure that the pressurizer level instrument drift evaluations were
 
4Q/2002 Inspection Findings - Indian Point 2                                                                                          Page 9 of 9 consistently bounded by the assumed instrument uncertainty within the safety analysis for a postulated Loss of Normal Feedwater event and a Loss of Offsite power event. The licensee documented this issue in condition report 2002000313.
Inspection Report# : 2002002(pdf)
Significance: N/A Feb 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO FOLLOW TAGGING PROCEDURE RESULTS IN INOPERABLE EDG An operator error during a tagout verification rendered the 21 emergency diesel generator (EDG) inoperable. This occurred when the 23 EDG was inoperable for planned maintenance. The tagout error was considered more than minor since it could reasonably be viewed as a precursor to a station blackout event and impacted mitigating systems cornerstone. The issue was determined to be of very low safety significance using phase two of the SDP because the exposure time was of very short duration (approximately five minutes), and the error was self-revealing so that operator action could be credited for timely restoration of the safety function. The failure to properly verify the tagout was a violation of TS 6.8.1.a. This is being treated as a Non-cited violation.
Inspection Report# : 2001014(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak. Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : March 25, 2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                    Page 1 of 11 Indian Point 2 1Q/2003 Plant Inspection Findings Initiating Events Significance:        Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements.
Green. A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements on August 13, 2002, which resulted in a steam generator level transient and required operator action to prevent a reactor trip. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event. Further, the controller replacement had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective of limiting conditions that affect plant stability. The finding was determined to be of very low safety significance (Green) because, although it affected stability of plant operating parameters, it did not increase the likelihood of a primary or secondary loss of coolant accident (LOCA), did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition.
Inspection Report# : 2003002(pdf)
Significance:        Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients.
Green. A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients during installation of a modification to replace nonsafety related steam generator system level controllers. The corrective actions from problems experienced during controller replacements on August 6 were ineffective to ensure that subsequent controllers replaced on August 9, August 13, and October 7, 2002 did not result in similar steam generator level transients and necessitate operator actions to prevent reactor trips. While this and the previous finding both concern problems with steam generator level replacements, the findings are distinct in that the previous finding identifies problems not entered into the corrective action program, while this finding concerns the ineffectiveness of corrective actions for problems that were entered into the corrective action program. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event, since the controller replacements had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective to limit conditions that challenge plant stability. However, the finding was similarly determined to be of very low safety significance (Green) because, although it affected stability of some plant parameters, it did not increase the likelihood of a primary or secondary LOCA, did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition Inspection Report# : 2003002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                    Page 2 of 11 Significance:        Aug 10, 2002 Identified By: Self Disclosing Item Type: FIN Finding CONTRACTOR WORKED OUTSIDE HIS ESTABLISHED JOB SCOPE FOR LANDSCAPING ACTIVITIES On July 19, 2002, a contractor worked outside his established job scope for landscaping activities. The consequences of this human performance error were the accidental electrocution of the individual and an offsite power electrical transient (loss of the 138 kilovolt station auxiliary transformer for approximately seven hours). This partial loss of offsite power event was more than minor, in that it impacted the reactor safety cornerstone with respect to the initiating event objective of limiting the likelihood of an event that upsets plant stability and challenges the critical safety function of the on-site emergency diesel generators. Notwithstanding the loss of life (which the Department of Labor, Occupational Safety and Health Administration is reviewing), this electrical transient event was of very low safety significance because it did not contribute to the likelihood of: loss of coolant accidents, a reactor trip and the unavailability of accident mitigation equipment or functions being unavailable; or of a fire or internal/external flood.
No violations of NRC requirements were identified.
Inspection Report# : 2002005(pdf)
Significance:        May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding REDUCTION OF PLANT POWER BY CONTROL ROOM OPERATORS DUE TO CONDENSATE PUMP MOTOR FAILURES On April 20, 2002, and on May 8, 2002, the control room operators reduced plant power due to condensate pump motor failures. A lack of a predictive maintenance program and an improperly set oil level indication system were the causes for two separate condensate motor failures. The events are more than minor since both events increased the likelihood of an initiating event. Operator response was necessary to ensure an automatic reactor trip did not occur due to a low steam generator level. The performance issues were of very low safety significance since there was no impact to normally available mitigating equipment.
Inspection Report# : 2002003(pdf)
Mitigating Systems Significance:        Dec 28, 2002 Identified By: NRC Item Type: FIN Finding UNTIMELY OPERABILITY DETERMINATION FOR THE 21 DIESEL GENERATOR On October 9, 2002, the licensee's organization did not identify in a timely manner that the 21 emergency diesel generator was inoperable. The causes for the untimely operability evaluation were fragmented communications between Entergy departments, untimely drip tank sample results, system engineering turnover, and a lack of sensitivity to a loss of the emergency power source safety function. The time between when the non-licensed operator had reported and added inventory to the jacket water expansion tank to the time the emergency diesel generator was declared inoperable was 7.5 hours which exceeded the limiting condition for operation within TS 3.0.1 to be in hot shutdown within seven hours. In the absence of reasonable expectation that a component is operable, the component shall be declared inoperable immediately. The untimely operability evaluation affects the mitigating systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                  Page 3 of 11 cornerstone objective. The attribute is human performance pre-event. This finding is of very low safety significance in phase 1 of the SDP since the 21 EDG was subsequently declared inoperable and actions within the TS were adhered to.
This finding did not result in an actual loss of the emergency on-site power source safety function nor did it increase the risk significance for external events. No violations of NRC requirements were identified.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE POST WORK TEST ON STEAM STOP CHECK VALVE The post work test on the 22 steam generator stop check valve (MS-41) failed to identify that the valve plug was installed upside down. This self-revealing event was identified on November 20, 2002, when operators responded to steam leak-by from this tagged closed valve that resulted in a fire alarm in the auxiliary feedwater pump room. This finding is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," in that, the post work test, PT-R67A, Reverse Flow Check at MS-41 and MS-42 Alternate Test, revision 1 did not adequately verify that MS-41 was properly reinstalled after preventative maintenance. The performance finding is considered more than minor, since the improperly installed valve plug would not have been identified prior to auxiliary feedwater system operability, had it not been identified during an unrelated tagout on the steam supply to the 22 auxiliary feedwater turbine. This is considered very low risk significance in accordance with NRC MC 0609 Appendix G since two alternate core cooling paths were available. This is an example of insufficient Entergy oversight of contractor work activities.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONFIGURATION CONTROL FOR A SAFETY RELATED SYSTEM The inspector identified an example of inadequate configuration control for a safety-related system. On November 20, 2002, the inspector identified that two 125 vdc circuit breakers were in their correct position (open) but administrative locking devices were not installed. The breakers are used to cross-connect the 21 and 22 125 vdc buses. This is considered a Non-Cited Violation of Technical Specification 6.8.1.a., which includes requirements for procedure adherence and operations of safety-related systems including the 125 volt DC system. Check off list (COL) 27.1.6, Instrument Buses, DC Distribution and PA Inverter, revision 18 requires the breakers to be open and locked. This performance deficiency is more than minor since more than one breaker was in the required position, but not locked.
The finding impacts the mitigating systems cornerstone and is associated with pre-event human error. The finding is considered very low safety significance (Green) since the operability or availability of the 21 and 22 DC buses were not impacted.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE CAUSE OF 23 EDG OUTPUT BREAKER TO CLOSE The inspector identified that Entergy did not adequately evaluate the cause blown control power fuses on the 23 EDG output breaker cubicle on November 10, 2002, that subsequently caused the breaker from closure on November 14, 2002. On November 14, 2002, Entergy identified the cause of the breaker failure as improper operation of the inertial latch mechanism. This performance issue is being treated as a Non-cited Violation of 10 CFR 50 Appendix B, Criterion file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                  Page 4 of 11 XVI. This violation is more than minor because the failure to identify the cause and preclude recurrence was considered a precursor to a more significant safety issue, in that, the plant could have started up with only 2 available EDGs - a violation of TS - and not have know it for approximately one month. The issue was determined to be of very low safety significance (Green) in accordance with MC 0609 Appendix G, since greater than three offsite and onsite power sources were available to cope with a postulated loss of offsite power.
Inspection Report# : 2002007(pdf)
Significance:        Sep 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 6.8 INVOLVING DEFICIENT GUIDANCE IN PROCEDURE AOI 27.1.1 Abnormal Operating Instruction (AOI) 27.1.1, "Loss of Normal Station Power," was deficient, in that no steps were provided in the procedure to identify that the lockout relays for the component cooling water (CCW) pumps were required to be reset following a loss and restoration of power to the motor supply breakers. This deficient procedure is being treated as a Non-Cited Violation of Technical Specification (TS) 6.8, "Procedures and Programs," in accordance with the NRC Enforcement Policy. The consequence of this finding was that the pump lockout relays would have prevented the 21 and 23 CCW pumps from starting automatically on low CCW system header pressure, for 12 days and 21 days, respectively. This finding represented a partial loss of the CCW system function and would reasonably have been corrected by operator action.
Inspection Report# : 2002006(pdf)
Significance:        Aug 10, 2002 Identified By: NRC Item Type: FIN Finding OPERATORS DID NOT IDENTIFY THE APPLICABILITY OF A SHUTDOWN TECHNICAL SPECIFICATION On July 19, 2002, operators did not identify the applicability of a shutdown Technical Specification (TS) associated with the planned removal from service of the 22 emergency diesel generator (EDG) while the 138 kilovolt off site power system was still out-of-service. This finding was associated with the reactor safety cornerstone with respect to the mitigating systems objective of ensuring the availability, reliability and capability of the EDG to respond to initiating events, such as a loss of offsite power, to prevent undesirable consequences. No violation of NRC requirements was identified, since Entergy restored the 22 EDG prior to exceeding the allowed outage time per TS 3.0.1. This finding was of very low safety significance since it did not represent a total loss of emergency power safety function.
Inspection Report# : 2002005(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902. The combined effect of the identified deficiencies was that, file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                      Page 5 of 11 as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program. Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls. [Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation TURBINE DRIVEN AUX FEED PUMP OIL ISSUES The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning three issues with the control and monitoring of lubrication oil used on the turbine driven auxiliary boiler feed water pump (22 ABFP). Each issue involved incomplete evaluations that led to repeat problems and potential for pump damage. The evaluation and corrective actions following identification in February 2002 that the wrong oil was added to the turbine speed governor were not fully effective. The evaluation of this issue identified that operators were not logging the quantity or specification of oil added during rounds or operation of equipment, but no actions were taken to address the issue. Additionally, the team noted that on July 10, while preparing to run the pump, Entergy identified additional confusion regarding the specification of oil to be added to the governor, an issue that should have been resolved.
Station personnel did not identify that oil analysis results in May 2002 showing a decrease in oil viscosity indicated that the wrong oil was likely added to a pump bearing and that corrective actions for a similar problem previously identified in May 2001 were ineffective. The evaluation and corrective actions following identification in October 2001 of issues with the required oil level in the pump inboard bearing were not fully effective, specifically the design drawing, the vendor manual, and operator training contained inconsistent information. These issues were evaluated using Phase I of the NRC SDP to have very low safety significance (Green), because pump operability was not directly affected. These issues are being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because the issues have been entered into Entergy's CAP. (NCV 50-247/02-010-002 Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SETPOINT DATABASE NOT CORRECTED FOR CIRCUIT BREAKER OVERCURRENT PROTECTION DEVICE SETPOINTS The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to promptly identify, determine the cause, and correct circuit breaker amptector setpoint database errors. The control of design setpoints is necessary to ensure the availability, reliability and capability of safety-related electrical systems. This issue was evaluated using Phase I of the NRC SDP and determined to have very low safety significance (Green), because the team did not identify any instances where a circuit breaker would not have been able to perform its safety function. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                  Page 6 of 11 (NCV 50-247/02-010-003)
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SAFETY INJECTION TOPPING PUMP VIBRATION CONSEQUENCES TO SAFETY-RELATED PIPING The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to identify that vibration of the non-safety-related SI accumulator topping pump caused stresses in adjacent safety-related piping that were above the code allowable values. The team evaluated this issue using Phase I of the NRC SDP, determining it to have very low safety significance (Green), because liquid penetrant examinations in the areas of high stress did not identify any piping damage. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP. (NCV 50-247/02-010-004 Inspection Report# : 2002010(pdf)
Significance:        Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding MULTIPLE GROUNDS ON THE PROTECTIVE CIRCUIT FOR UNIT 1 SUBSTATION 102NS3 RESULTED IN A LOSS OF THE 13.8 KILOVOLT LIGHTING AND POWER BUS SECTION 3 On May 17, 2002, multiple grounds on the protective circuit for Unit 1 substation 102NS3 resulted in a loss of the 13.8 kilovolt (kv) lighting and power bus section 3. The consequence of this event was a loss of alternate safe shutdown power to all major alternate safe shutdown pumps and selected instrumentation. At the time, the Unit 2 normal and emergency electrical power supplies were available to supply power to the above stated mitigation equipment and instrumentation. The licensee repaired and restored the 13.8 kv bus section 3 within 30 hours of the fault. The performance issue is inadequate retirement of protective circuits for 440 volt substations (132PC3 and 142PC3) that could impact availability of alternate safe shutdown power supplies. This issue is more than minor since unavailability of alternate safe shutdown equipment for 30 hours is viewed as a precursor to a significant event and the alternate safe shutdown power supplies are a risk-significant maintenance rule system which was unavailable for greater than 24 hours.
Inspection Report# : 2002004(pdf)
Significance:        Nov 05, 2001 Identified By: NRC Item Type: FIN Finding CREW HIGH FAILURE RATE DURING THE 2001 ANNUAL REQUALIFICATION SIMULATOR EXAMINATIONS This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with crew high failure rate (four of seven crews failed) during facility-administered annual licensed operator requalification examinations conducted last fall. The finding was previously characterized as having substantial safety significance (Yellow) in NRC Inspection Report 50-247/01-13. The inspectors noted that the licensee's evaluation identified a fundamental underlying weakness: The station has yet to overcome cultural weaknesses that include an unwillingness to confront poor performance, an over reliance on procedures to change behavior, and compartmentalization. More specifically, the licensee identified three root causes: 1) Operations training had not focused on the basic building blocks that ensure a healthy program; 2) The station had not maintained a core of career oriented, plant knowledgeable instructors and operators; and 3) Operations department involvement with Operations Training had often been file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                  Page 7 of 11 ineffective. The inspectors concluded that the methodology and level of detail of the licensee's root cause evaluation were reasonable. The licensee implemented a number of corrective actions to address the identified causes. The corrective actions are described in the station's Training Improvement Plan. The more significant corrective actions included initiatives that aimed to 1) improve the quality of training and training materials; 2) increase the number of instructors who have Unit 2 plant experience; and 3) provide additional management support and oversight of training.
The inspectors determined that the corrective actions are appropriately focused on the identified causes. These actions were appropriately prioritized, and either complete or scheduled for completion. Notably, the licensee took strong immediate corrective actions following the requalification examination failures to provide extensive retraining to each shift, and continue to provide this high intensity training. The inspectors independently assessed the extent of the underlying conditions that led to the Yellow finding and found that performance issues had also existed in other Operations Training programs, such as initial licensed operator and non-licensed operator training programs. These problems existed for at least three years, both prior to and following the steam generator tube failure event in 2001.
Although licensee audits and assessments had identified most of the performance problems prior to the crew high failure rate, they did not identify long-term operator performance as a concern. The inspectors concluded that the licensee's extent of condition review appropriately bounded the underlying conditions that led to the Yellow finding as evidenced by the fact that the licensee had also identified the duration and extent of the problems, and the failure to recognize the long standing issues. (Updated) FIN 05000247/01-013-01: Proposed finding due to crew high failure rate during the 2001 annual requalification simulator examinations. This finding was documented in an October 2001 inspection and initially characterized as a potential Yellow finding, the final safety significance to be determined (TBD). This finding was subsequently evaluated under the significance determination process (SDP) and characterized as (reference NRC to Entergy letters dated December 5, 2001, and February 28, 2002). The 95002 Supplemental Inspection (reference Inspection Report No. 50-247/02-09, dated May 31, 2002), assessed the licensee's evaluation of the crew high failure rates and the corrective actions taken to address this performance issue. As stated in the cover letter to Inspection Report No. 50-247/02-09, this finding remains open until after the completion of Entergy's licensed operator requalification examinations, scheduled for September-October 2002, and further review by the NRC. This item remains open.
Inspection Report# : 2002009(pdf)
Inspection Report# : 2001013(pdf)
Inspection Report# : 2002004(pdf)
Inspection Report# : 2003008(pdf)
Barrier Integrity Significance:        Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Ineffective tracking of Inservice Testing of component cooling water system relief valve.
The team identified a finding regarding the scheduled inservice test (IST) of a Component Cooling Water (CCW) system pressure relief valve that was inadvertently not performed during the last plant refueling outage. Access and testing of this valve normally requires the plant to be shutdown. This finding was not a violation of applicable technical specifications for IST because the timing of the team's questions identified this issue to Entergy personnel while the test was within the allowed test interval extension. However, the issue was more than minor because, if left uncorrected, improperly tracked relief valve tests could result in a more significant safety concern because the valves would not be tested as required to ensure their reliable operation to provide CCW piping over-pressurization protection during accident conditions and maintain the CCW containment penetration barrier integrity. The finding was determined to be of very low safety significance (Green) because there was no actual open pathway in the physical containment structure. (Section 1R21b. 1.1) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                Page 8 of 11 Inspection Report# : 2003004(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation OPERATORS DEVIATE FROM PLANT OPERATING PROCEDURES On November 23, 2002, during a plant cooldown, Entergy deviated from the guidance of plant operating procedure (POP) 3.3, Plant Cooldown, Rev. 57. The consequence of the failure to follow the POP guidance was to exceed the operational limits on the steam generator tube sheet differential pressure of 1600 psid with a maximum value of approximately 1855 psid. Control room operators were unaware of this operational limit. Reviews of steam generator manufacturer specifications and the Updated Final Safety Analysis Report design basis accident analysis information indicated that the steam generator tubes were designed to withstand up to 2485 psid during upset and hydrostatic conditions. Therefore, the structural integrity and qualification of the steam generator tubes was maintained. Failure to document the basis of marking non-conditional steps in POP 3.3 as not-applicable is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V. This issue was considered more than minor because it represented a lack of understanding of procedure requirements and awareness of plant operational limitations. This finding is considered very low safety significance (Green) in accordance with manual chapter 0609 Appendix G, in that the core cooling pathway via the steam generators was not impacted.
Inspection Report# : 2002007(pdf)
Significance:      Jun 29, 2002 Identified By: NRC Item Type: FIN Finding DURING SURVEILLANCE TESTING OF THE SAFETY INJECTION DISCHARGE MOTOR-OPERATED VALVE 851B, THE VALVE FAILED TO STROKE CLOSED On May 27, 2002, during surveillance testing of the safety injection discharge motor-operated valve (851B), the valve failed to stroke closed. The initial operability evaluation did not consider the non-automatic containment isolation function for this valve. This event was documented in condition report No. 200205433. The performance issue associated with this finding is a weakness in operator knowledge of multi-function safety system components. This is the second recent example where operators did not consider this function for a safety-related valve. The first example was documented in NRC report 50-247/2002-003, section 1R15. The untimely and incomplete operability assessment for safety injection discharge valve 851B has very low safety significance since the containment isolation valve was restored to an operable status prior to exceeding Technical Specification 3.6.A.3.a.2.d limiting condition for operation.
Inspection Report# : 2002004(pdf)
Significance:      May 11, 2002 Identified By: Licensee Item Type: FIN Finding UNTIMELY OPERATOR EVALUATION FOR CONTAINMENT ISOLATION VALVE 869B On April 11, 2002, operators did not complete a timely operability evaluation for containment isolation valve 869B after the disconnect switch operating handle on motor control center (MCC)26BB broke while applying an equipment tagout. At the time, the operators neither verified that the disconnect would operate nor completed an adequate evaluation regarding the ability to close valve 869B to perform its containment isolation function. An operability evaluation was completed about six hours later by a different operating crew and the operators then entered a four-hour limiting condition for operation and isolated the containment penetration per the technical specifications 3.6.A.3.a.2.b.
The untimely operability evaluation increased the unavailability time for the containment spray system. The inoperable containment isolation valve issue was more than minor because it impacts the containment barrier. This issue had very file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                    Page 9 of 11 low safety significance since the containment isolation valve was repaired and restored to an operable status prior to exceeding technical specification 3.6.A.3.a.2.d. This issue was an example of untimely operator implementation of technical specification requirements in response to degraded safety equipment.
Inspection Report# : 2002003(pdf)
Emergency Preparedness Significance:      Sep 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECT PREVIOUSLY IDENTIFIED CONDITION IN THE JNC REGARDING THE TIMELY AND ACCURATE DISSEMINATION OF INFORMATION During the emergency plan exercise conducted on September 24, 2002, the licensee JNC personnel proceeded with a 1:55 p.m. press conference informing the media that no release was in progress when, according to the exercise events, a release had begun just prior to the press conference. This information was not in conflict with the emergency alert system message that was in effect at the time of the briefing. However, the failure to provide updated information could cause confusion for those receiving it through the media. This failure was a previously identified weakness as documented in Drill Critique Reports and in condition reports. This is a non-cited violation of 10 CFR 50 Appendix E Section IV.F.2.g which requires, in part, that weaknesses or deficiencies that are identified during a drill or exercise shall be corrected. The risk associated with this release of incorrect information was determined to be of very low significance because it does not constitute a loss of function in meeting the applicable planning standard (10 CFR 50.47 (b)(14).
Inspection Report# : 2002012(pdf)
Significance: N/A May 11, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF 10 CFR 50.54(q) FOR ACCOUNTABILITY On March 6, 2002, the licensee implemented changes to the accountability process that decreased the effectiveness of the Emergency Plan (E-Plan). The finding was considered more than minor because, if left uncorrected, it would become a more significant safety concern. Changing commitments in the E-Plan without prior approval impacts the NRC's ability to perform its regulatory function and potentially creates an ineffective response to a radiological emergency. The consequences of this change were minimal because it did not preclude the function of accountability from being performed, albeit delayed. The licensee has implemented the corrective actions and has since met the timeliness goal. This change which decreased the effectiveness of the Plan is being treateed as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued May 1, 2000.
Inspection Report# : 2002003(pdf)
Occupational Radiation Safety Public Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                                Page 10 of 11 Physical Protection Significance:      Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation UNIT 2 SECURITY RESPONSE FORCE MEMBER WAS FOUND INATTENTIVE TO ASSIGNED DUTIES On July 29, 2002, a member of the Unit 2 security response force was found inattentive to assigned duties. This inspector identified finding was treated as a non-cited violation of 10 CFR 73.55(b)(1)(i), and the Indian Point 2 Physical Security Plan. The security response force officer's inattentiveness to duties was determined to have very low safety significance, using the Interim Physical Significance Determination Process. The finding did not involve a significant compromise of the Physical Security Plan; no actual intrusion occurred; and, there have not been greater than two similar findings in the past four quarters.
Inspection Report# : 2002005(pdf)
Miscellaneous Significance: N/A Feb 03, 2003 Identified By: NRC Item Type: FIN Finding Generally effective Corrrective Action Program implementation The inspectors concluded that, within the scope of the issues reviewed, overall, Indian Point 2 (IP2) personnel were identifying issues at a threshold suitable to recognize conditions adverse to quality and help ensure reliable equipment operation. Although station backlogs (corrective actions, maintenance and engineering items) remained relatively high, the inspectors observed that senior management continued to provide reasonable oversight and emphasis on accountability for corrective action program performance. Corrective action process condition reports adequately characterized and bounded the scope of the problems, and correctly assessed equipment operability. Nevertheless, the team identified instances regarding steam generator level controller replacement problems and a cable tunnel groundwater leak where problems were not identified and not entered into the corrective action process. IP2 personnel usually evaluated problems to a level of detail appropriate to its technical complexity and risk significance. Problems were adequately prioritized for resolution considering the potential safety significance of the issues and their probability for recurrence. However, in some instances (emergency diesel generator wiring termination and breaker setpoint database), the inspectors identified evaluations where the problems were not completely addressed. Corrective actions generally addressed the problems and encompassed the scope of the issues. Based on the issues reviewed, the inspectors found corrective actions were scheduled and completed commensurate with the risk significance of the issues. Formal effectiveness reviews were completed, and then reviewed by the Corrective Action Review Board to help ensure the corrective actions were effective in resolving more significant problems. Notwithstanding, corrective actions were not effective to prevent repetitive problems during a steam generator controller replacement modification.
Inspection Report# : 2003002(pdf)
Significance:      Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAIL TO USE THE APPROPRIATE TOOLING DEVICE FOR MOVEMENT OF FUEL ASSEMBLY G28 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Indian Point 2                                                              Page 11 of 11 ON JULY 23, 2002 GREEN. On July 23, 2002, Entergy did not appropriately evaluate and implement short-term actions associated with Condition Report (CR) IP2-2002-07253. The consequence of the finding was the relocation of spent fuel assembly G-28 without the appropriate handling tools and precautions. The finding is more than minor since it could be reasonably viewed as a precursor to a significant event (dropped spent fuel assembly in the spent fuel pool). The Significance Determination Process is not modeled for a finding of this type. However, in accordance with NRC Manual Chapter 0612, this finding was reviewed by NRC risk analysts and management and has been determined to be of very low safety significance because no actual consequence existed and there was no unintended radiation worker exposure. The finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion V, and is being treated as a non-cited violation. (1R20)
Inspection Report# : 2002005(pdf)
Significance: N/A Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Improved performance in the areas of design control, equipment and human performance, and corrective actions.
Overall, the team found that Entergy operated IP2 safely and that through implementation of the FIP, progress had been made in improving performance in the areas of design control, equipment and human performance, and corrective actions, Specifically, the team determined overall: quality of engineering products has improved, and design and licensing basis control have been strengthened; equipment performance improved (including reduced backlogs of corrective maintenance work orders, operator workarounds, CCR deficiencies, etc); station human error rate and number of equipment mis-positioning events have declined; and, improved effectiveness of corrective action program, including identification and documentation of issues at a low threshold.
Inspection Report# : 2002010(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : June 17, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          07/22/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                    Page 1 of 11 Indian Point 2 2Q/2003 Plant Inspection Findings Initiating Events Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER EMERGENT WORK PACKAGE INSTRUCTIONS FOR 22 STEAM GENERATOR LEVEL BISTABLE REPLACEMENT On February 7, 2003, a self-revealing finding involved inadequate emergent work instructions that resulted in an electrical short during replacement of the 22 steam generator low level bistable. The electrical short caused a breaker trip on circuit 10 of instrument bus 21and the resultant loss of electrical power to the pressurizer level and reactor coolant system pressure control channels (failed low). The inadequate work instructions is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V, since the instructions did not account for consideration of performing this replacement with the circuit de-energized or the proximity to other reactor protection system relays. The performance issue is more than minor since the operators were required to take action to restore reactor coolant system pressure and pressurizer level to preclude a reactor trip. The finding involves the initiating events cornerstone in that it increased the likelihood of upset in plant stability and it involves human error during the planning of an emergent work activity. This finding is considered to be of very low safety significance in that in accordance with NRC Manual Chapter 0609, Appendix A, the finding did not contribute to the likelihood of a secondary or primary LOCA initiator and it did not contribute to either a reactor trip or mitigation system unavailability.
Inspection Report# : 2003003(pdf)
Significance:      Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements.
Green. A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements on August 13, 2002, which resulted in a steam generator level transient and required operator action to prevent a reactor trip. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event. Further, the controller replacement had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective of limiting conditions that affect plant stability. The finding was determined to be of very low safety significance (Green) because, although it affected stability of plant operating parameters, it did not increase the likelihood of a primary or secondary loss of coolant accident (LOCA), did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition.
Inspection Report# : 2003002(pdf)
Significance:      Feb 03, 2003 Identified By: NRC Item Type: FIN Finding file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                    Page 2 of 11 A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients.
Green. A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients during installation of a modification to replace nonsafety related steam generator system level controllers. The corrective actions from problems experienced during controller replacements on August 6 were ineffective to ensure that subsequent controllers replaced on August 9, August 13, and October 7, 2002 did not result in similar steam generator level transients and necessitate operator actions to prevent reactor trips. While this and the previous finding both concern problems with steam generator level replacements, the findings are distinct in that the previous finding identifies problems not entered into the corrective action program, while this finding concerns the ineffectiveness of corrective actions for problems that were entered into the corrective action program. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event, since the controller replacements had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective to limit conditions that challenge plant stability. However, the finding was similarly determined to be of very low safety significance (Green) because, although it affected stability of some plant parameters, it did not increase the likelihood of a primary or secondary LOCA, did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition Inspection Report# : 2003002(pdf)
Significance:        Aug 10, 2002 Identified By: Self Disclosing Item Type: FIN Finding CONTRACTOR WORKED OUTSIDE HIS ESTABLISHED JOB SCOPE FOR LANDSCAPING ACTIVITIES On July 19, 2002, a contractor worked outside his established job scope for landscaping activities. The consequences of this human performance error were the accidental electrocution of the individual and an offsite power electrical transient (loss of the 138 kilovolt station auxiliary transformer for approximately seven hours). This partial loss of offsite power event was more than minor, in that it impacted the reactor safety cornerstone with respect to the initiating event objective of limiting the likelihood of an event that upsets plant stability and challenges the critical safety function of the on-site emergency diesel generators. Notwithstanding the loss of life (which the Department of Labor, Occupational Safety and Health Administration is reviewing), this electrical transient event was of very low safety significance because it did not contribute to the likelihood of: loss of coolant accidents, a reactor trip and the unavailability of accident mitigation equipment or functions being unavailable; or of a fire or internal/external flood.
No violations of NRC requirements were identified.
Inspection Report# : 2002005(pdf)
Mitigating Systems Significance:        Jun 28, 2003 Identified By: NRC Item Type: FIN Finding INADEQUATE OPERABILITY EVALUATION FOR THE 13.8 KV SYSTEM The inspector identified that the licensee's operability evaluation during a 13.8 KV system reduced voltage test was inadequate. The operability evaluation did not evaluate accident load carrying capability as defined in the technical specification basis and it did not address communications and protocols between the distribution company and the file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                      Page 3 of 11 licensee to restore from the test in a timely manner. NRC Manual Chapter 9900 states that when a system's capability is degraded to a point where it cannot perform with reasonable assurance of reliability, the system should be judged inoperable. The finding was more than minor because it impacted the attribute of the mitigating system cornerstone objective. Specifically, the cornerstone objective is to ensure that the 13.8 KV system is capable of performing its safety function during a postulated loss of normal power event without undesirable consequences. This finding was determined to be of low safety significance because it did not result in the actual loss of the offsite power supply safety function.
Inspection Report# : 2003007(pdf)
Significance:      Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions associated with an unauthorized modification to the No. 22 component cooling water pump.
The inspector identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI. The licensee did not evaluate and take effective corrective actions associated with a material substitution for the 22 component cooling water pump inboard bearing oil level indication system. The bearing oil level indication system contributed to the failure of the #22 CCW pump on December 5, 2002. This finding is greater than minor since it is associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective. The inspectors conducted a Phase 1 SDP screening and determined that the failure to take effective corrective action on #22 CCW pump was of a very low safety significance since the redundant train components were operable and unaffected by this unauthorized modification.
Inspection Report# : 2003007(pdf)
Significance:      May 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH THE 23 EDG LOAD SWINGS BETWEEN MAY 2000 AND FEBRUARY 2003 The inspectors identified that ineffective corrective actions resulted in repetitive surveillance test failures of the 23 emergency diesel generator between December 2001 and February 2003. This finding is considered a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI. The finding is more than minor because the surveillance test failures impacted the availability of one train of emergency AC power source. This finding was of very low risk significance because the repetitive failures did not result in an actual loss of function for the emergency AC power.
Inspection Report# : 2003003(pdf)
Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation POST-WORK TEST INADEQUAATE FOR 22 BORIC ACID TRANSFER PUMP BORIC ACID FILTER STOP VALVE A self-revealing event was identified on February 26, 2003, when operators observed no boric acid flow to the reactor vessel via the No. 22 boric acid transfer pump (BATP). It was determined that during preventative maintenance activities in March 2001, the post-work test on the No. 22 BATP outlet valve to the boric acid filter stop was inadequate to identify that the valve finger plate was installed upside down. This finding is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V. This event is considered more than minor because the improperly installed valve plate affected the availability of one train of emergency boration. This is considered to be of very low file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                  Page 4 of 11 risk significance in accordance with NRC MC 0609 Appendix A, since the emergency boration function was not lost due to this performance issue.
Inspection Report# : 2003003(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: FIN Finding UNTIMELY OPERABILITY DETERMINATION FOR THE 21 DIESEL GENERATOR On October 9, 2002, the licensee's organization did not identify in a timely manner that the 21 emergency diesel generator was inoperable. The causes for the untimely operability evaluation were fragmented communications between Entergy departments, untimely drip tank sample results, system engineering turnover, and a lack of sensitivity to a loss of the emergency power source safety function. The time between when the non-licensed operator had reported and added inventory to the jacket water expansion tank to the time the emergency diesel generator was declared inoperable was 7.5 hours which exceeded the limiting condition for operation within TS 3.0.1 to be in hot shutdown within seven hours. In the absence of reasonable expectation that a component is operable, the component shall be declared inoperable immediately. The untimely operability evaluation affects the mitigating systems cornerstone objective. The attribute is human performance pre-event. This finding is of very low safety significance in phase 1 of the SDP since the 21 EDG was subsequently declared inoperable and actions within the TS were adhered to.
This finding did not result in an actual loss of the emergency on-site power source safety function nor did it increase the risk significance for external events. No violations of NRC requirements were identified.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE CAUSE OF 23 EDG OUTPUT BREAKER TO CLOSE The inspector identified that Entergy did not adequately evaluate the cause blown control power fuses on the 23 EDG output breaker cubicle on November 10, 2002, that subsequently caused the breaker from closure on November 14, 2002. On November 14, 2002, Entergy identified the cause of the breaker failure as improper operation of the inertial latch mechanism. This performance issue is being treated as a Non-cited Violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is more than minor because the failure to identify the cause and preclude recurrence was considered a precursor to a more significant safety issue, in that, the plant could have started up with only 2 available EDGs - a violation of TS - and not have know it for approximately one month. The issue was determined to be of very low safety significance (Green) in accordance with MC 0609 Appendix G, since greater than three offsite and onsite power sources were available to cope with a postulated loss of offsite power.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE POST WORK TEST ON STEAM STOP CHECK VALVE The post work test on the 22 steam generator stop check valve (MS-41) failed to identify that the valve plug was installed upside down. This self-revealing event was identified on November 20, 2002, when operators responded to steam leak-by from this tagged closed valve that resulted in a fire alarm in the auxiliary feedwater pump room. This finding is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," in that, the post work test, PT-R67A, Reverse Flow Check at MS-41 and MS-42 Alternate Test, revision 1 did not adequately verify that MS-41 was properly reinstalled after preventative maintenance. The performance finding file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                  Page 5 of 11 is considered more than minor, since the improperly installed valve plug would not have been identified prior to auxiliary feedwater system operability, had it not been identified during an unrelated tagout on the steam supply to the 22 auxiliary feedwater turbine. This is considered very low risk significance in accordance with NRC MC 0609 Appendix G since two alternate core cooling paths were available. This is an example of insufficient Entergy oversight of contractor work activities.
Inspection Report# : 2002007(pdf)
Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONFIGURATION CONTROL FOR A SAFETY RELATED SYSTEM The inspector identified an example of inadequate configuration control for a safety-related system. On November 20, 2002, the inspector identified that two 125 vdc circuit breakers were in their correct position (open) but administrative locking devices were not installed. The breakers are used to cross-connect the 21 and 22 125 vdc buses. This is considered a Non-Cited Violation of Technical Specification 6.8.1.a., which includes requirements for procedure adherence and operations of safety-related systems including the 125 volt DC system. Check off list (COL) 27.1.6, Instrument Buses, DC Distribution and PA Inverter, revision 18 requires the breakers to be open and locked. This performance deficiency is more than minor since more than one breaker was in the required position, but not locked.
The finding impacts the mitigating systems cornerstone and is associated with pre-event human error. The finding is considered very low safety significance (Green) since the operability or availability of the 21 and 22 DC buses were not impacted.
Inspection Report# : 2002007(pdf)
Significance:        Sep 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 6.8 INVOLVING DEFICIENT GUIDANCE IN PROCEDURE AOI 27.1.1 Abnormal Operating Instruction (AOI) 27.1.1, "Loss of Normal Station Power," was deficient, in that no steps were provided in the procedure to identify that the lockout relays for the component cooling water (CCW) pumps were required to be reset following a loss and restoration of power to the motor supply breakers. This deficient procedure is being treated as a Non-Cited Violation of Technical Specification (TS) 6.8, "Procedures and Programs," in accordance with the NRC Enforcement Policy. The consequence of this finding was that the pump lockout relays would have prevented the 21 and 23 CCW pumps from starting automatically on low CCW system header pressure, for 12 days and 21 days, respectively. This finding represented a partial loss of the CCW system function and would reasonably have been corrected by operator action.
Inspection Report# : 2002006(pdf)
Significance:        Aug 10, 2002 Identified By: NRC Item Type: FIN Finding OPERATORS DID NOT IDENTIFY THE APPLICABILITY OF A SHUTDOWN TECHNICAL SPECIFICATION On July 19, 2002, operators did not identify the applicability of a shutdown Technical Specification (TS) associated with the planned removal from service of the 22 emergency diesel generator (EDG) while the 138 kilovolt off site power system was still out-of-service. This finding was associated with the reactor safety cornerstone with respect to the mitigating systems objective of ensuring the availability, reliability and capability of the EDG to respond to initiating events, such as a loss of offsite power, to prevent undesirable consequences. No violation of NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                    Page 6 of 11 requirements was identified, since Entergy restored the 22 EDG prior to exceeding the allowed outage time per TS 3.0.1. This finding was of very low safety significance since it did not represent a total loss of emergency power safety function.
Inspection Report# : 2002005(pdf)
Significance:      Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902. The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program. Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls. [Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Significance:      Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation TURBINE DRIVEN AUX FEED PUMP OIL ISSUES The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning three issues with the control and monitoring of lubrication oil used on the turbine driven auxiliary boiler feed water pump (22 ABFP). Each issue involved incomplete evaluations that led to repeat problems and potential for pump damage. The evaluation and corrective actions following identification in February 2002 that the wrong oil was added to the turbine speed governor were not fully effective. The evaluation of this issue identified that operators were not logging the quantity or specification of oil added during rounds or operation of equipment, but no actions were taken to address the issue. Additionally, the team noted that on July 10, while preparing to run the pump, Entergy identified additional confusion regarding the specification of oil to be added to the governor, an issue that should have been resolved.
Station personnel did not identify that oil analysis results in May 2002 showing a decrease in oil viscosity indicated that the wrong oil was likely added to a pump bearing and that corrective actions for a similar problem previously identified in May 2001 were ineffective. The evaluation and corrective actions following identification in October 2001 of issues with the required oil level in the pump inboard bearing were not fully effective, specifically the design drawing, the vendor manual, and operator training contained inconsistent information. These issues were evaluated using Phase I of the NRC SDP to have very low safety significance (Green), because pump operability was not directly affected. These issues are being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                      Page 7 of 11 on the very low safety significance, and because the issues have been entered into Entergy's CAP. (NCV 50-247/02-010-002 Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SETPOINT DATABASE NOT CORRECTED FOR CIRCUIT BREAKER OVERCURRENT PROTECTION DEVICE SETPOINTS The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to promptly identify, determine the cause, and correct circuit breaker amptector setpoint database errors. The control of design setpoints is necessary to ensure the availability, reliability and capability of safety-related electrical systems. This issue was evaluated using Phase I of the NRC SDP and determined to have very low safety significance (Green), because the team did not identify any instances where a circuit breaker would not have been able to perform its safety function. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP.
(NCV 50-247/02-010-003)
Inspection Report# : 2002010(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation SAFETY INJECTION TOPPING PUMP VIBRATION CONSEQUENCES TO SAFETY-RELATED PIPING The team identified a non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action concerning the failure to identify that vibration of the non-safety-related SI accumulator topping pump caused stresses in adjacent safety-related piping that were above the code allowable values. The team evaluated this issue using Phase I of the NRC SDP, determining it to have very low safety significance (Green), because liquid penetrant examinations in the areas of high stress did not identify any piping damage. This issue is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy based on the very low safety significance, and because it has been entered into Entergy's CAP. (NCV 50-247/02-010-004 Inspection Report# : 2002010(pdf)
Barrier Integrity Significance:        Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Ineffective tracking of Inservice Testing of component cooling water system relief valve.
The team identified a finding regarding the scheduled inservice test (IST) of a Component Cooling Water (CCW) system pressure relief valve that was inadvertently not performed during the last plant refueling outage. Access and testing of this valve normally requires the plant to be shutdown. This finding was not a violation of applicable technical specifications for IST because the timing of the team's questions identified this issue to Entergy personnel while the test was within the allowed test interval extension. However, the issue was more than minor because, if left uncorrected, improperly tracked relief valve tests could result in a more significant safety concern because the valves would not be file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                    Page 8 of 11 tested as required to ensure their reliable operation to provide CCW piping over-pressurization protection during accident conditions and maintain the CCW containment penetration barrier integrity. The finding was determined to be of very low safety significance (Green) because there was no actual open pathway in the physical containment structure. (Section 1R21b. 1.1)
Inspection Report# : 2003004(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation OPERATORS DEVIATE FROM PLANT OPERATING PROCEDURES On November 23, 2002, during a plant cooldown, Entergy deviated from the guidance of plant operating procedure (POP) 3.3, Plant Cooldown, Rev. 57. The consequence of the failure to follow the POP guidance was to exceed the operational limits on the steam generator tube sheet differential pressure of 1600 psid with a maximum value of approximately 1855 psid. Control room operators were unaware of this operational limit. Reviews of steam generator manufacturer specifications and the Updated Final Safety Analysis Report design basis accident analysis information indicated that the steam generator tubes were designed to withstand up to 2485 psid during upset and hydrostatic conditions. Therefore, the structural integrity and qualification of the steam generator tubes was maintained. Failure to document the basis of marking non-conditional steps in POP 3.3 as not-applicable is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V. This issue was considered more than minor because it represented a lack of understanding of procedure requirements and awareness of plant operational limitations. This finding is considered very low safety significance (Green) in accordance with manual chapter 0609 Appendix G, in that the core cooling pathway via the steam generators was not impacted.
Inspection Report# : 2002007(pdf)
Emergency Preparedness Significance:      Sep 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECT PREVIOUSLY IDENTIFIED CONDITION IN THE JNC REGARDING THE TIMELY AND ACCURATE DISSEMINATION OF INFORMATION During the emergency plan exercise conducted on September 24, 2002, the licensee JNC personnel proceeded with a 1:55 p.m. press conference informing the media that no release was in progress when, according to the exercise events, a release had begun just prior to the press conference. This information was not in conflict with the emergency alert system message that was in effect at the time of the briefing. However, the failure to provide updated information could cause confusion for those receiving it through the media. This failure was a previously identified weakness as documented in Drill Critique Reports and in condition reports. This is a non-cited violation of 10 CFR 50 Appendix E Section IV.F.2.g which requires, in part, that weaknesses or deficiencies that are identified during a drill or exercise shall be corrected. The risk associated with this release of incorrect information was determined to be of very low significance because it does not constitute a loss of function in meeting the applicable planning standard (10 CFR 50.47 (b)(14).
Inspection Report# : 2002012(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                    Page 9 of 11 Occupational Radiation Safety Public Radiation Safety Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH PACKAGING PROCEDURES A self-revealing non-cited violation of 10 CFR 71.12 was identified for failure to comply with shipping cask package procedures. On February 6, 2003, a CNS 8-120 B cask was received from the Indian Point Energy Center at a consolidation facility in South Carolina with a bolt missing on the primary lid's pressure test port in violation of the cask use and maintenance procedures. This finding was more than minor in that it was associated with the Public Radiation Safety Cornerstone's attribute of procedures for transportation packages. The finding affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials contained in an NRC-approved Type B package released into the public domain. The finding was determined to be of very low safety significance in that the finding did not involve exceeding transportation radiation limits, there was no breach of the package during transit, and the issue was a Certificate of Compliance maintenance/use performance deficiency.
Inspection Report# : 2003003(pdf)
Physical Protection Significance:      Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation UNIT 2 SECURITY RESPONSE FORCE MEMBER WAS FOUND INATTENTIVE TO ASSIGNED DUTIES On July 29, 2002, a member of the Unit 2 security response force was found inattentive to assigned duties. This inspector identified finding was treated as a non-cited violation of 10 CFR 73.55(b)(1)(i), and the Indian Point 2 Physical Security Plan. The security response force officer's inattentiveness to duties was determined to have very low safety significance, using the Interim Physical Significance Determination Process. The finding did not involve a significant compromise of the Physical Security Plan; no actual intrusion occurred; and, there have not been greater than two similar findings in the past four quarters.
Inspection Report# : 2002005(pdf)
Miscellaneous Significance: N/A Feb 03, 2003 Identified By: NRC Item Type: FIN Finding Generally effective Corrrective Action Program implementation file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                                Page 10 of 11 The inspectors concluded that, within the scope of the issues reviewed, overall, Indian Point 2 (IP2) personnel were identifying issues at a threshold suitable to recognize conditions adverse to quality and help ensure reliable equipment operation. Although station backlogs (corrective actions, maintenance and engineering items) remained relatively high, the inspectors observed that senior management continued to provide reasonable oversight and emphasis on accountability for corrective action program performance. Corrective action process condition reports adequately characterized and bounded the scope of the problems, and correctly assessed equipment operability. Nevertheless, the team identified instances regarding steam generator level controller replacement problems and a cable tunnel groundwater leak where problems were not identified and not entered into the corrective action process. IP2 personnel usually evaluated problems to a level of detail appropriate to its technical complexity and risk significance. Problems were adequately prioritized for resolution considering the potential safety significance of the issues and their probability for recurrence. However, in some instances (emergency diesel generator wiring termination and breaker setpoint database), the inspectors identified evaluations where the problems were not completely addressed. Corrective actions generally addressed the problems and encompassed the scope of the issues. Based on the issues reviewed, the inspectors found corrective actions were scheduled and completed commensurate with the risk significance of the issues. Formal effectiveness reviews were completed, and then reviewed by the Corrective Action Review Board to help ensure the corrective actions were effective in resolving more significant problems. Notwithstanding, corrective actions were not effective to prevent repetitive problems during a steam generator controller replacement modification.
Inspection Report# : 2003002(pdf)
Significance:      Aug 10, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAIL TO USE THE APPROPRIATE TOOLING DEVICE FOR MOVEMENT OF FUEL ASSEMBLY G28 ON JULY 23, 2002 GREEN. On July 23, 2002, Entergy did not appropriately evaluate and implement short-term actions associated with Condition Report (CR) IP2-2002-07253. The consequence of the finding was the relocation of spent fuel assembly G-28 without the appropriate handling tools and precautions. The finding is more than minor since it could be reasonably viewed as a precursor to a significant event (dropped spent fuel assembly in the spent fuel pool). The Significance Determination Process is not modeled for a finding of this type. However, in accordance with NRC Manual Chapter 0612, this finding was reviewed by NRC risk analysts and management and has been determined to be of very low safety significance because no actual consequence existed and there was no unintended radiation worker exposure. The finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion V, and is being treated as a non-cited violation. (1R20)
Inspection Report# : 2002005(pdf)
Significance: N/A Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Improved performance in the areas of design control, equipment and human performance, and corrective actions.
Overall, the team found that Entergy operated IP2 safely and that through implementation of the FIP, progress had been made in improving performance in the areas of design control, equipment and human performance, and corrective actions, Specifically, the team determined overall: quality of engineering products has improved, and design and licensing basis control have been strengthened; equipment performance improved (including reduced backlogs of corrective maintenance work orders, operator workarounds, CCR deficiencies, etc); station human error rate and number of equipment mis-positioning events have declined; and, improved effectiveness of corrective action program, including identification and documentation of issues at a low threshold.
Inspection Report# : 2002010(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Indian Point 2                                                              Page 11 of 11 Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          10/08/2003
 
3Q/2003 Inspection Findings - Indian Point 2                                                                      Page 1 of 9 Indian Point 2 3Q/2003 Plant Inspection Findings Initiating Events Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER EMERGENT WORK PACKAGE INSTRUCTIONS FOR 22 STEAM GENERATOR LEVEL BISTABLE REPLACEMENT On February 7, 2003, a self-revealing finding involved inadequate emergent work instructions that resulted in an electrical short during replacement of the 22 steam generator low level bistable. The electrical short caused a breaker trip on circuit 10 of instrument bus 21and the resultant loss of electrical power to the pressurizer level and reactor coolant system pressure control channels (failed low). The inadequate work instructions is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V, since the instructions did not account for consideration of performing this replacement with the circuit de-energized or the proximity to other reactor protection system relays.
The performance issue is more than minor since the operators were required to take action to restore reactor coolant system pressure and pressurizer level to preclude a reactor trip. The finding involves the initiating events cornerstone in that it increased the likelihood of upset in plant stability and it involves human error during the planning of an emergent work activity. This finding is considered to be of very low safety significance in that in accordance with NRC Manual Chapter 0609, Appendix A, the finding did not contribute to the likelihood of a secondary or primary LOCA initiator and it did not contribute to either a reactor trip or mitigation system unavailability.
Inspection Report# : 2003003(pdf)
Significance:      Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements.
Green. A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements on August 13, 2002, which resulted in a steam generator level transient and required operator action to prevent a reactor trip.
This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event.
Further, the controller replacement had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective of limiting conditions that affect plant stability. The finding was determined to be of very low safety significance (Green) because, although it affected stability of plant operating parameters, it did not increase the likelihood of a primary or secondary loss of coolant accident (LOCA), did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition.
Inspection Report# : 2003002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                      Page 2 of 9 Significance:        Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients.
Green. A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients during installation of a modification to replace nonsafety related steam generator system level controllers.
The corrective actions from problems experienced during controller replacements on August 6 were ineffective to ensure that subsequent controllers replaced on August 9, August 13, and October 7, 2002 did not result in similar steam generator level transients and necessitate operator actions to prevent reactor trips. While this and the previous finding both concern problems with steam generator level replacements, the findings are distinct in that the previous finding identifies problems not entered into the corrective action program, while this finding concerns the ineffectiveness of corrective actions for problems that were entered into the corrective action program. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event, since the controller replacements had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective to limit conditions that challenge plant stability.
However, the finding was similarly determined to be of very low safety significance (Green) because, although it affected stability of some plant parameters, it did not increase the likelihood of a primary or secondary LOCA, did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition Inspection Report# : 2003002(pdf)
Mitigating Systems Significance:        Sep 27, 2003 Identified By: NRC Item Type: FIN Finding THE PERFORMANCE FINDING INVOLVED INADEQUATE SHORT TERM CORRECTIVE ACTIONS ASSOCIATED WITH FIRE LEAKS ON A FIRE HEADER IN THE UNIT 1 TURBINE BUILDING The inspectors identified a finding involving inadequate corrective actions associated with multiple leaks on a six-inch fire header in the Unit 1 turbine building. On September 10, 2003, an 80 gallon per minute fire header leak occurred that operators isolated by depressurizing the entire fire water suppression system at Unit 2 for approximately three hours. This leak occurred approximately one foot from a similar through-wall leak which occurred on July 16, 2003.
This performance issue is considered more than minor based on example 4.f. in MC 0612 Appendix E. The performance finding involves the Mitigating Systems Cornerstone objective of fire suppression system availability to respond to fires. The finding is very low risk significance based upon the results from the fire protection risk significance screening methodology (FPRSSM). The finding impacts both manual suppression capability and automatic suppression capability.
Inspection Report# : 2003011(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                      Page 3 of 9 Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B, CRITERION V. THE PROCEDURAL STEPS FOR THE INSTALLATION OF A FLEXIBLE COUPLING WERE NOT ADEQUATE TO VERIFY THAT THE COMPONENT WAS PROPERLY INSTALLED.
The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V. In November 2002, a maintenance work instruction to install a 21 emergency diesel generator (EDG) service water supply flexible coupling did not include critical installation steps per the vendor manual. This resulted in a significant service water leak from the expansion joint on August 14, 2003.
This finding is greater than minor since if left uncorrected, it could be a more significant safety concern as this type of flexible coupling is used on all three EDGs. The inspectors determined that the expansion joint leakage was of a very low safety significance since it did not adversely impact service water cooling to the emergency diesel generator or the overall service water system cooling capability, did not impact equipment and functions associated with internal flooding in the diesel generator room, and did not result in a loss of service water or emergency power safety function that contributed to internal flooding initiated events.
Inspection Report# : 2003011(pdf)
Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B. A DESIGN CHANGE PACKAGE DID NOT ACCURATELY REFLECT ACTUAL PLANT CONDITIONS AND RESULTED IN AN UNINTENDED PLANT TRANSIENT The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, involving the design change package (DCP-200105716-I) to replace a pressurizer level recorder which did not contain accurate design details. As a consequence, during installation of the design change an unintended plant transient challenged operators.
This finding is greater than minor based upon NRC Manual Chapter 0612, Appendix E, example 4.b. This finding is of very low safety significance. The finding did contribute to the likelihood of a reactor trip; however, it did not impact the availability of mitigation equipment, increase the likelihood of a primary or secondary system LOCA, or increase the likelihood of an internal fire or flood.
Inspection Report# : 2003011(pdf)
Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation EQUIPMENT TAGOUT TO RESTORE THE 22 SEAL INJECTION FILTER WERE INADEQUATE TO MAINTAIN PROPER CONFIGURATION CONTROL OF THE SYSTEM The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, involving an incomplete procedure for restoring to service the 22 seal injection filter from maintenance. The consequence was an approximate 70 gallon per minute chemical volume and control system leak through an open vent valve which lasted for approximately two minutes before operators identified and shut the vent valve.
This finding is more than minor since it adversely impacted the Mitigating System Cornerstone objective of safety system capability and availability with respect to the attributes of configuration control and procedural quality. The inadequate restoration procedure resulted in a significant chemical and volume control system leak (the capacity of one file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                        Page 4 of 9 coolant charging pump) that degraded normal charging flow and emergency boration capability for a short period of time. The finding is of very low safety significance since it did not result in a loss of emergency boration safety function.
Inspection Report# : 2003011(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: FIN Finding INADEQUATE OPERABILITY EVALUATION FOR THE 13.8 KV SYSTEM The inspector identified that the licensee's operability evaluation during a 13.8 KV system reduced voltage test was inadequate. The operability evaluation did not evaluate accident load carrying capability as defined in the technical specification basis and it did not address communications and protocols between the distribution company and the licensee to restore from the test in a timely manner. NRC Manual Chapter 9900 states that when a system's capability is degraded to a point where it cannot perform with reasonable assurance of reliability, the system should be judged inoperable.
The finding was more than minor because it impacted the attribute of the mitigating system cornerstone objective.
Specifically, the cornerstone objective is to ensure that the 13.8 KV system is capable of performing its safety function during a postulated loss of normal power event without undesirable consequences. This finding was determined to be of low safety significance because it did not result in the actual loss of the offsite power supply safety function.
Inspection Report# : 2003007(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions associated with an unauthorized modification to the No. 22 component cooling water pump.
The inspector identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI. The licensee did not evaluate and take effective corrective actions associated with a material substitution for the 22 component cooling water pump inboard bearing oil level indication system. The bearing oil level indication system contributed to the failure of the #22 CCW pump on December 5, 2002.
This finding is greater than minor since it is associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective. The inspectors conducted a Phase 1 SDP screening and determined that the failure to take effective corrective action on #22 CCW pump was of a very low safety significance since the redundant train components were operable and unaffected by this unauthorized modification.
Inspection Report# : 2003007(pdf)
Significance:        May 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH THE 23 EDG LOAD SWINGS BETWEEN MAY 2000 AND FEBRUARY 2003 The inspectors identified that ineffective corrective actions resulted in repetitive surveillance test failures of the 23 emergency diesel generator between December 2001 and February 2003. This finding is considered a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI. The finding is more than minor because the surveillance test failures impacted the availability of one train of emergency AC power source. This finding was of very low risk file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                    Page 5 of 9 significance because the repetitive failures did not result in an actual loss of function for the emergency AC power.
Inspection Report# : 2003003(pdf)
Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation POST-WORK TEST INADEQUAATE FOR 22 BORIC ACID TRANSFER PUMP BORIC ACID FILTER STOP VALVE A self-revealing event was identified on February 26, 2003, when operators observed no boric acid flow to the reactor vessel via the No. 22 boric acid transfer pump (BATP). It was determined that during preventative maintenance activities in March 2001, the post-work test on the No. 22 BATP outlet valve to the boric acid filter stop was inadequate to identify that the valve finger plate was installed upside down. This finding is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V. This event is considered more than minor because the improperly installed valve plate affected the availability of one train of emergency boration. This is considered to be of very low risk significance in accordance with NRC MC 0609 Appendix A, since the emergency boration function was not lost due to this performance issue.
Inspection Report# : 2003003(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: FIN Finding UNTIMELY OPERABILITY DETERMINATION FOR THE 21 DIESEL GENERATOR On October 9, 2002, the licensee's organization did not identify in a timely manner that the 21 emergency diesel generator was inoperable. The causes for the untimely operability evaluation were fragmented communications between Entergy departments, untimely drip tank sample results, system engineering turnover, and a lack of sensitivity to a loss of the emergency power source safety function. The time between when the non-licensed operator had reported and added inventory to the jacket water expansion tank to the time the emergency diesel generator was declared inoperable was 7.5 hours which exceeded the limiting condition for operation within TS 3.0.1 to be in hot shutdown within seven hours. In the absence of reasonable expectation that a component is operable, the component shall be declared inoperable immediately. The untimely operability evaluation affects the mitigating systems cornerstone objective. The attribute is human performance pre-event. This finding is of very low safety significance in phase 1 of the SDP since the 21 EDG was subsequently declared inoperable and actions within the TS were adhered to.
This finding did not result in an actual loss of the emergency on-site power source safety function nor did it increase the risk significance for external events. No violations of NRC requirements were identified.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE CAUSE OF 23 EDG OUTPUT BREAKER TO CLOSE The inspector identified that Entergy did not adequately evaluate the cause blown control power fuses on the 23 EDG output breaker cubicle on November 10, 2002, that subsequently caused the breaker from closure on November 14, 2002. On November 14, 2002, Entergy identified the cause of the breaker failure as improper operation of the inertial latch mechanism. This performance issue is being treated as a Non-cited Violation of 10 CFR 50 Appendix B, Criterion XVI. This violation is more than minor because the failure to identify the cause and preclude recurrence was considered a precursor to a more significant safety issue, in that, the plant could have started up with only 2 available EDGs - a violation of TS - and not have know it for approximately one month. The issue was determined to be of very file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                  Page 6 of 9 low safety significance (Green) in accordance with MC 0609 Appendix G, since greater than three offsite and onsite power sources were available to cope with a postulated loss of offsite power.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE POST WORK TEST ON STEAM STOP CHECK VALVE The post work test on the 22 steam generator stop check valve (MS-41) failed to identify that the valve plug was installed upside down. This self-revealing event was identified on November 20, 2002, when operators responded to steam leak-by from this tagged closed valve that resulted in a fire alarm in the auxiliary feedwater pump room. This finding is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings," in that, the post work test, PT-R67A, Reverse Flow Check at MS-41 and MS-42 Alternate Test, revision 1 did not adequately verify that MS-41 was properly reinstalled after preventative maintenance. The performance finding is considered more than minor, since the improperly installed valve plug would not have been identified prior to auxiliary feedwater system operability, had it not been identified during an unrelated tagout on the steam supply to the 22 auxiliary feedwater turbine. This is considered very low risk significance in accordance with NRC MC 0609 Appendix G since two alternate core cooling paths were available. This is an example of insufficient Entergy oversight of contractor work activities.
Inspection Report# : 2002007(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONFIGURATION CONTROL FOR A SAFETY RELATED SYSTEM The inspector identified an example of inadequate configuration control for a safety-related system. On November 20, 2002, the inspector identified that two 125 vdc circuit breakers were in their correct position (open) but administrative locking devices were not installed. The breakers are used to cross-connect the 21 and 22 125 vdc buses. This is considered a Non-Cited Violation of Technical Specification 6.8.1.a., which includes requirements for procedure adherence and operations of safety-related systems including the 125 volt DC system. Check off list (COL) 27.1.6, Instrument Buses, DC Distribution and PA Inverter, revision 18 requires the breakers to be open and locked. This performance deficiency is more than minor since more than one breaker was in the required position, but not locked.
The finding impacts the mitigating systems cornerstone and is associated with pre-event human error. The finding is considered very low safety significance (Green) since the operability or availability of the 21 and 22 DC buses were not impacted.
Inspection Report# : 2002007(pdf)
Significance:      Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                    Page 7 of 9 found not to be constructed in accordance with UL U902.
The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program.
Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls.
[Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Barrier Integrity Significance:      Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Ineffective tracking of Inservice Testing of component cooling water system relief valve.
The team identified a finding regarding the scheduled inservice test (IST) of a Component Cooling Water (CCW) system pressure relief valve that was inadvertently not performed during the last plant refueling outage. Access and testing of this valve normally requires the plant to be shutdown.
This finding was not a violation of applicable technical specifications for IST because the timing of the team's questions identified this issue to Entergy personnel while the test was within the allowed test interval extension.
However, the issue was more than minor because, if left uncorrected, improperly tracked relief valve tests could result in a more significant safety concern because the valves would not be tested as required to ensure their reliable operation to provide CCW piping over-pressurization protection during accident conditions and maintain the CCW containment penetration barrier integrity. The finding was determined to be of very low safety significance (Green) because there was no actual open pathway in the physical containment structure. (Section 1R21b. 1.1)
Inspection Report# : 2003004(pdf)
Significance:      Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation OPERATORS DEVIATE FROM PLANT OPERATING PROCEDURES On November 23, 2002, during a plant cooldown, Entergy deviated from the guidance of plant operating procedure (POP) 3.3, Plant Cooldown, Rev. 57. The consequence of the failure to follow the POP guidance was to exceed the operational limits on the steam generator tube sheet differential pressure of 1600 psid with a maximum value of file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                    Page 8 of 9 approximately 1855 psid. Control room operators were unaware of this operational limit. Reviews of steam generator manufacturer specifications and the Updated Final Safety Analysis Report design basis accident analysis information indicated that the steam generator tubes were designed to withstand up to 2485 psid during upset and hydrostatic conditions. Therefore, the structural integrity and qualification of the steam generator tubes was maintained. Failure to document the basis of marking non-conditional steps in POP 3.3 as not-applicable is considered a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V. This issue was considered more than minor because it represented a lack of understanding of procedure requirements and awareness of plant operational limitations. This finding is considered very low safety significance (Green) in accordance with manual chapter 0609 Appendix G, in that the core cooling pathway via the steam generators was not impacted.
Inspection Report# : 2002007(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH PACKAGING PROCEDURES A self-revealing non-cited violation of 10 CFR 71.12 was identified for failure to comply with shipping cask package procedures. On February 6, 2003, a CNS 8-120 B cask was received from the Indian Point Energy Center at a consolidation facility in South Carolina with a bolt missing on the primary lid's pressure test port in violation of the cask use and maintenance procedures. This finding was more than minor in that it was associated with the Public Radiation Safety Cornerstone's attribute of procedures for transportation packages. The finding affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials contained in an NRC-approved Type B package released into the public domain. The finding was determined to be of very low safety significance in that the finding did not involve exceeding transportation radiation limits, there was no breach of the package during transit, and the issue was a Certificate of Compliance maintenance/use performance deficiency.
Inspection Report# : 2003003(pdf)
Physical Protection Miscellaneous file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Indian Point 2                                                                    Page 9 of 9 Significance: N/A Feb 03, 2003 Identified By: NRC Item Type: FIN Finding Generally effective Corrrective Action Program implementation The inspectors concluded that, within the scope of the issues reviewed, overall, Indian Point 2 (IP2) personnel were identifying issues at a threshold suitable to recognize conditions adverse to quality and help ensure reliable equipment operation. Although station backlogs (corrective actions, maintenance and engineering items) remained relatively high, the inspectors observed that senior management continued to provide reasonable oversight and emphasis on accountability for corrective action program performance. Corrective action process condition reports adequately characterized and bounded the scope of the problems, and correctly assessed equipment operability. Nevertheless, the team identified instances regarding steam generator level controller replacement problems and a cable tunnel groundwater leak where problems were not identified and not entered into the corrective action process.
IP2 personnel usually evaluated problems to a level of detail appropriate to its technical complexity and risk significance. Problems were adequately prioritized for resolution considering the potential safety significance of the issues and their probability for recurrence. However, in some instances (emergency diesel generator wiring termination and breaker setpoint database), the inspectors identified evaluations where the problems were not completely addressed.
Corrective actions generally addressed the problems and encompassed the scope of the issues. Based on the issues reviewed, the inspectors found corrective actions were scheduled and completed commensurate with the risk significance of the issues. Formal effectiveness reviews were completed, and then reviewed by the Corrective Action Review Board to help ensure the corrective actions were effective in resolving more significant problems.
Notwithstanding, corrective actions were not effective to prevent repetitive problems during a steam generator controller replacement modification.
Inspection Report# : 2003002(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            01/12/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                  Page 1 of 10 Indian Point 2 4Q/2003 Plant Inspection Findings Initiating Events Significance:      Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE APPROPRIATE AND TIMELY CORRECTIVE ACTIONS TO ADDRESS THE REPEATED GRID-RELATED REACTOR TRIPS OF UNIT 2 This team-identified finding involves inadequate corrective actions for repeat Unit 2 reactor scrams attributed to grid-related faults and associated protective relaying failures. The lack of thorough evaluations and corrective actions on the part of Entergy, in cooperation with the responsible Transmission and Distribution Operator for the local area electrical grid, have resulted in an increased frequency of plant transients and consequential challenges to Unit 2 safety related systems and licensed operators.
This finding is greater than minor because it affects the Initiating Events Cornerstone and represents an increased likelihood of an event that challenges critical safety functions and operator response. Using the Indian Point Unit 2 Significance Determination Process Phase 2 "Transient with Power Conversion System Available" worksheet, this finding was determined to be of very low safety significance.
Inspection Report# : 2003013(pdf)
Significance:      Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation TS 6.8.1 VIOLATION - FAILURE TO ADHERE TO EMERGENCY OPERATING PROCEDURE ES-0.1, CONTINUOUS ACTION STEP 1.0 ON AUGUST 3, 2003 The team identified a violation involving the failure of an operating crew to adhere to a continuous action step of Emergency Operating Procedure ES-0.1, "Reactor Trip Response," resulting in an avoidable plant transient.
Specifically, in response to the reactor trip and partial loss of offsite power (LOOP) event on August 3, 2003, the Unit 2 operating crew did not correctly implement continuous action step 1 of ES-0.1, which led to the cycling of the pressurizer power-operated relief valves (PORVs) ten times, complicating reactor coolant system (RCS) pressure control.
This finding is greater than minor because it affected the Initiating Events Cornerstone and could reasonably be viewed as a precursor to a more significant event, in that, the failure to implement established procedures could place the reactor outside its design envelope and, for this particular event, the repeated cycling of the PORVs could have resulted in a loss of coolant event had a PORV stuck open. This finding is of very low safety significance because all mitigation systems were available during the event and was treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2003013(pdf)
Significance:      May 15, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                    Page 2 of 10 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER EMERGENT WORK PACKAGE INSTRUCTIONS FOR 22 STEAM GENERATOR LEVEL BISTABLE REPLACEMENT On February 7, 2003, a self-revealing finding involved inadequate emergent work instructions that resulted in an electrical short during replacement of the 22 steam generator low level bistable. The electrical short caused a breaker trip on circuit 10 of instrument bus 21and the resultant loss of electrical power to the pressurizer level and reactor coolant system pressure control channels (failed low). The inadequate work instructions is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V, since the instructions did not account for consideration of performing this replacement with the circuit de-energized or the proximity to other reactor protection system relays.
The performance issue is more than minor since the operators were required to take action to restore reactor coolant system pressure and pressurizer level to preclude a reactor trip. The finding involves the initiating events cornerstone in that it increased the likelihood of upset in plant stability and it involves human error during the planning of an emergent work activity. This finding is considered to be of very low safety significance in that in accordance with NRC Manual Chapter 0609, Appendix A, the finding did not contribute to the likelihood of a secondary or primary LOCA initiator and it did not contribute to either a reactor trip or mitigation system unavailability.
Inspection Report# : 2003003(pdf)
Significance:      Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements.
Green. A failure to initiate a condition report to identify problems associated with nonsafety related steam generator level controller replacements on August 13, 2002, which resulted in a steam generator level transient and required operator action to prevent a reactor trip.
This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event.
Further, the controller replacement had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective of limiting conditions that affect plant stability. The finding was determined to be of very low safety significance (Green) because, although it affected stability of plant operating parameters, it did not increase the likelihood of a primary or secondary loss of coolant accident (LOCA), did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition.
Inspection Report# : 2003002(pdf)
Significance:      Feb 03, 2003 Identified By: NRC Item Type: FIN Finding A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients.
Green. A self-revealing finding was identified for ineffective corrective actions to prevent main feedwater flow and steam generator level transients during installation of a modification to replace nonsafety related steam generator system level controllers.
The corrective actions from problems experienced during controller replacements on August 6 were ineffective to ensure that subsequent controllers replaced on August 9, August 13, and October 7, 2002 did not result in similar steam file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                    Page 3 of 10 generator level transients and necessitate operator actions to prevent reactor trips. While this and the previous finding both concern problems with steam generator level replacements, the findings are distinct in that the previous finding identifies problems not entered into the corrective action program, while this finding concerns the ineffectiveness of corrective actions for problems that were entered into the corrective action program. This issue is more than minor because the problem could reasonably be viewed as a precursor to a significant event, since the controller replacements had an actual impact on feedwater flow and steam generator level control which required operator action to preclude a reactor trip. This issue affects the initiating event cornerstone objective to limit conditions that challenge plant stability.
However, the finding was similarly determined to be of very low safety significance (Green) because, although it affected stability of some plant parameters, it did not increase the likelihood of a primary or secondary LOCA, did not contribute to a reactor trip and a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flooding condition Inspection Report# : 2003002(pdf)
Mitigating Systems Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation A Green NCV was identified for failure to take appropriate corrective actions for Gas Turbine 1 An NCV of 10 CFR 50.65 (a)(1) was identified when Entergy failed to take appropriate corrective actions when the #1 Gas Turbine (GT1) exceeded its maintenance rule (a)(1) reliability monitoring goal. This finding was greater than minor because it affected the reliability of GT1 which is used to mitigate the consequences of a station blackout. This issue was evaluated using the significance determination process and determined to be of very low significance (Green) since the redundant train was always available to perform the GT safety functions.
Inspection Report# : 2003012(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: FIN Finding THE PERFORMANCE FINDING INVOLVED INADEQUATE SHORT TERM CORRECTIVE ACTIONS ASSOCIATED WITH FIRE LEAKS ON A FIRE HEADER IN THE UNIT 1 TURBINE BUILDING The inspectors identified a finding involving inadequate corrective actions associated with multiple leaks on a six-inch fire header in the Unit 1 turbine building. On September 10, 2003, an 80 gallon per minute fire header leak occurred that operators isolated by depressurizing the entire fire water suppression system at Unit 2 for approximately three hours. This leak occurred approximately one foot from a similar through-wall leak which occurred on July 16, 2003.
This performance issue is considered more than minor based on example 4.f. in MC 0612 Appendix E. The performance finding involves the Mitigating Systems Cornerstone objective of fire suppression system availability to respond to fires. The finding is very low risk significance based upon the results from the fire protection risk significance screening methodology (FPRSSM). The finding impacts both manual suppression capability and automatic suppression capability.
Inspection Report# : 2003011(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                    Page 4 of 10 Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B, CRITERION V. THE PROCEDURAL STEPS FOR THE INSTALLATION OF A FLEXIBLE COUPLING WERE NOT ADEQUATE TO VERIFY THAT THE COMPONENT WAS PROPERLY INSTALLED.
The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V. In November 2002, a maintenance work instruction to install a 21 emergency diesel generator (EDG) service water supply flexible coupling did not include critical installation steps per the vendor manual. This resulted in a significant service water leak from the expansion joint on August 14, 2003.
This finding is greater than minor since if left uncorrected, it could be a more significant safety concern as this type of flexible coupling is used on all three EDGs. The inspectors determined that the expansion joint leakage was of a very low safety significance since it did not adversely impact service water cooling to the emergency diesel generator or the overall service water system cooling capability, did not impact equipment and functions associated with internal flooding in the diesel generator room, and did not result in a loss of service water or emergency power safety function that contributed to internal flooding initiated events.
Inspection Report# : 2003011(pdf)
Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B. A DESIGN CHANGE PACKAGE DID NOT ACCURATELY REFLECT ACTUAL PLANT CONDITIONS AND RESULTED IN AN UNINTENDED PLANT TRANSIENT The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, involving the design change package (DCP-200105716-I) to replace a pressurizer level recorder which did not contain accurate design details. As a consequence, during installation of the design change an unintended plant transient challenged operators.
This finding is greater than minor based upon NRC Manual Chapter 0612, Appendix E, example 4.b. This finding is of very low safety significance. The finding did contribute to the likelihood of a reactor trip; however, it did not impact the availability of mitigation equipment, increase the likelihood of a primary or secondary system LOCA, or increase the likelihood of an internal fire or flood.
Inspection Report# : 2003011(pdf)
Significance:      Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation EQUIPMENT TAGOUT TO RESTORE THE 22 SEAL INJECTION FILTER WERE INADEQUATE TO MAINTAIN PROPER CONFIGURATION CONTROL OF THE SYSTEM The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, involving an incomplete procedure for restoring to service the 22 seal injection filter from maintenance. The consequence was an approximate 70 gallon per minute chemical volume and control system leak through an open vent valve which lasted for approximately two minutes before operators identified and shut the vent valve.
This finding is more than minor since it adversely impacted the Mitigating System Cornerstone objective of safety system capability and availability with respect to the attributes of configuration control and procedural quality. The inadequate restoration procedure resulted in a significant chemical and volume control system leak (the capacity of one file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                      Page 5 of 10 coolant charging pump) that degraded normal charging flow and emergency boration capability for a short period of time. The finding is of very low safety significance since it did not result in a loss of emergency boration safety function.
Inspection Report# : 2003011(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: FIN Finding INADEQUATE OPERABILITY EVALUATION FOR THE 13.8 KV SYSTEM The inspector identified that the licensee's operability evaluation during a 13.8 KV system reduced voltage test was inadequate. The operability evaluation did not evaluate accident load carrying capability as defined in the technical specification basis and it did not address communications and protocols between the distribution company and the licensee to restore from the test in a timely manner. NRC Manual Chapter 9900 states that when a system's capability is degraded to a point where it cannot perform with reasonable assurance of reliability, the system should be judged inoperable.
The finding was more than minor because it impacted the attribute of the mitigating system cornerstone objective.
Specifically, the cornerstone objective is to ensure that the 13.8 KV system is capable of performing its safety function during a postulated loss of normal power event without undesirable consequences. This finding was determined to be of low safety significance because it did not result in the actual loss of the offsite power supply safety function.
Inspection Report# : 2003007(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions associated with an unauthorized modification to the No. 22 component cooling water pump.
The inspector identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI. The licensee did not evaluate and take effective corrective actions associated with a material substitution for the 22 component cooling water pump inboard bearing oil level indication system. The bearing oil level indication system contributed to the failure of the #22 CCW pump on December 5, 2002.
This finding is greater than minor since it is associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective. The inspectors conducted a Phase 1 SDP screening and determined that the failure to take effective corrective action on #22 CCW pump was of a very low safety significance since the redundant train components were operable and unaffected by this unauthorized modification.
Inspection Report# : 2003007(pdf)
Significance:        May 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH THE 23 EDG LOAD SWINGS BETWEEN MAY 2000 AND FEBRUARY 2003 The inspectors identified that ineffective corrective actions resulted in repetitive surveillance test failures of the 23 emergency diesel generator between December 2001 and February 2003. This finding is considered a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI. The finding is more than minor because the surveillance test failures impacted the availability of one train of emergency AC power source. This finding was of very low risk file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                              04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                    Page 6 of 10 significance because the repetitive failures did not result in an actual loss of function for the emergency AC power.
Inspection Report# : 2003003(pdf)
Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation POST-WORK TEST INADEQUAATE FOR 22 BORIC ACID TRANSFER PUMP BORIC ACID FILTER STOP VALVE A self-revealing event was identified on February 26, 2003, when operators observed no boric acid flow to the reactor vessel via the No. 22 boric acid transfer pump (BATP). It was determined that during preventative maintenance activities in March 2001, the post-work test on the No. 22 BATP outlet valve to the boric acid filter stop was inadequate to identify that the valve finger plate was installed upside down. This finding is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V. This event is considered more than minor because the improperly installed valve plate affected the availability of one train of emergency boration. This is considered to be of very low risk significance in accordance with NRC MC 0609 Appendix A, since the emergency boration function was not lost due to this performance issue.
Inspection Report# : 2003003(pdf)
Significance:      Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902.
The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program.
Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls.
[Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Inspection Report# : 2004003(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                  Page 7 of 10 Barrier Integrity Significance:      Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Ineffective tracking of Inservice Testing of component cooling water system relief valve.
The team identified a finding regarding the scheduled inservice test (IST) of a Component Cooling Water (CCW) system pressure relief valve that was inadvertently not performed during the last plant refueling outage. Access and testing of this valve normally requires the plant to be shutdown.
This finding was not a violation of applicable technical specifications for IST because the timing of the team's questions identified this issue to Entergy personnel while the test was within the allowed test interval extension.
However, the issue was more than minor because, if left uncorrected, improperly tracked relief valve tests could result in a more significant safety concern because the valves would not be tested as required to ensure their reliable operation to provide CCW piping over-pressurization protection during accident conditions and maintain the CCW containment penetration barrier integrity. The finding was determined to be of very low safety significance (Green) because there was no actual open pathway in the physical containment structure. (Section 1R21b. 1.1)
Inspection Report# : 2003004(pdf)
Emergency Preparedness Significance:      Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to implement appropriate corrective actions for degraded Technical Support Center Batteries The inspectors identified a findings of very low safety significance (Green) regarding the licensee's failure to implement appropriate corrective actions for degraded TSC batteries. The perofrmance deficiency associated with this findings was the failure to take timely and effective corrective actions for the degraded TSC batteries. The degraded batteries adversely impacted the Non-Risk Significant Planning Standard, as described in 10 CFR 50.47(b)(8), which requires adequate facilities and equipment be maintained to support emergency response. This finding was more than minor significance because the batteries were allowed to remain in a degraded stated in excess of twenty-four hours without adequate measures to ensure that their TSC support function would be maintained. The finding is of very low safety significance, because the subsequent analysis indicated that the battery banks remained functional in this condition.
Inspection Report# : 2004003(pdf)
Significance:      Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions for repetitive failures of the plant vent noble gas effluent monitor The team identified a finding of very low safety significance (Green) regarding the licensee's inadequate corrective actions for repetitive failures of a TS required surveillance of the plant vent noble gas effluent monitor. Since July file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                  Page 8 of 10 2002, the monitor has failed the surveillance test five of the six times performed. The performance deficiency associated with this findings was inadequate corrective actions for repetitive failures of a TS required surveillance. This finding is more than minor significance because the R-27 radiation monitor was removed from service for periods in excess of twenty-four hours as a result of inadequate corrective actions. The finding was evaluated using the EP SDP, and was screened to be of very low safety significance, because there were alternate monitoring methods available in the event of an accident.
Inspection Report# : 2004003(pdf)
Significance:      Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE EOF UPSs ON AUGUST 14 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the August 14, 2003, loss of off-site power event which revealed that Entergy did not have a preventive maintenance program in place to ensure the continued functionality of the numerous un-interruptible power supplies in the Emergency Operations Facility (EOF) which provide back-up power to emergency response equipment.
This finding is considered greater than minor because a significant amount of the Unit 2 and Unit 3 emergency response organization communications equipment was non-functional on August 14 until off-site power was restored.
However, this finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions from the EOF, TSC/OSC, and Unit 2 and 3 central control rooms, using back-up telephone communications.
Inspection Report# : 2003013(pdf)
Significance:      Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE UNIT 2 TSC DIESEL ON AUGUST 14, 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the failure of the Unit 2 TSC back-up diesel generator to function on August 14, 2003. The conditions which caused the diesel generator to fail to function involved electrical loading of the diesel generator in excess of its design capacity. This condition was initially identified in February 2000 and not resolved in a timely manner.
This finding is considered more than minor because a significant amount of TSC/OSC emergency response equipment, necessary to implement the Emergency Plan, was either de-energized by the Entergy staff because of the loss of sufficient air conditioning to ensure emergency response equipment would not be damaged due to overheating, or was without AC power because the diesel was non-functional. This finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions.
Inspection Report# : 2003013(pdf)
Occupational Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                    Page 9 of 10 Public Radiation Safety Significance:      May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH PACKAGING PROCEDURES A self-revealing non-cited violation of 10 CFR 71.12 was identified for failure to comply with shipping cask package procedures. On February 6, 2003, a CNS 8-120 B cask was received from the Indian Point Energy Center at a consolidation facility in South Carolina with a bolt missing on the primary lid's pressure test port in violation of the cask use and maintenance procedures. This finding was more than minor in that it was associated with the Public Radiation Safety Cornerstone's attribute of procedures for transportation packages. The finding affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials contained in an NRC-approved Type B package released into the public domain. The finding was determined to be of very low safety significance in that the finding did not involve exceeding transportation radiation limits, there was no breach of the package during transit, and the issue was a Certificate of Compliance maintenance/use performance deficiency.
Inspection Report# : 2003003(pdf)
Physical Protection Miscellaneous Significance: N/A Dec 12, 2003 Identified By: NRC Item Type: FIN Finding The licensee's Corrective Action Program, used for identifying, tracking, prioritizing, and resolving deficiencies, was appropriately implemented in most cases.
The inspection team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program, evaluating and prioritizing issues, and implementing appropriate corrective actions.
However, the inspectors identified some minor equipment problems that had not been identified in the corrective action program. The evaluation of problems wa generally adequate, but the inspectors identified two Green findings related to the failure to implement effective corrective acitons for degraded emergency preparedness equipment. These findings were determined not to involve violations of NRC requirements, however, they represented additional examples of the substantive cross-cutting issue in the problem identification and resolution area that was identified during the previous assessments. Based on interviews conducted during the inspection, station personnel felt free to identify safety issues and enter them into the corrective action program.
Inspection Report# : 2004003(pdf)
Significance: N/A Feb 03, 2003 Identified By: NRC Item Type: FIN Finding Generally effective Corrrective Action Program implementation file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Indian Point 2                                                                Page 10 of 10 The inspectors concluded that, within the scope of the issues reviewed, overall, Indian Point 2 (IP2) personnel were identifying issues at a threshold suitable to recognize conditions adverse to quality and help ensure reliable equipment operation. Although station backlogs (corrective actions, maintenance and engineering items) remained relatively high, the inspectors observed that senior management continued to provide reasonable oversight and emphasis on accountability for corrective action program performance. Corrective action process condition reports adequately characterized and bounded the scope of the problems, and correctly assessed equipment operability. Nevertheless, the team identified instances regarding steam generator level controller replacement problems and a cable tunnel groundwater leak where problems were not identified and not entered into the corrective action process.
IP2 personnel usually evaluated problems to a level of detail appropriate to its technical complexity and risk significance. Problems were adequately prioritized for resolution considering the potential safety significance of the issues and their probability for recurrence. However, in some instances (emergency diesel generator wiring termination and breaker setpoint database), the inspectors identified evaluations where the problems were not completely addressed.
Corrective actions generally addressed the problems and encompassed the scope of the issues. Based on the issues reviewed, the inspectors found corrective actions were scheduled and completed commensurate with the risk significance of the issues. Formal effectiveness reviews were completed, and then reviewed by the Corrective Action Review Board to help ensure the corrective actions were effective in resolving more significant problems.
Notwithstanding, corrective actions were not effective to prevent repetitive problems during a steam generator controller replacement modification.
Inspection Report# : 2003002(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak.
Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\IP2\ip2_pim.html                                                            04/22/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 1 of 7 Indian Point 2 1Q/2004 Plant Inspection Findings Initiating Events Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE APPROPRIATE AND TIMELY CORRECTIVE ACTIONS TO ADDRESS THE REPEATED GRID-RELATED REACTOR TRIPS OF UNIT 2 This team-identified finding involves inadequate corrective actions for repeat Unit 2 reactor scrams attributed to grid-related faults and associated protective relaying failures. The lack of thorough evaluations and corrective actions on the part of Entergy, in cooperation with the responsible Transmission and Distribution Operator for the local area electrical grid, have resulted in an increased frequency of plant transients and consequential challenges to Unit 2 safety related systems and licensed operators.
This finding is greater than minor because it affects the Initiating Events Cornerstone and represents an increased likelihood of an event that challenges critical safety functions and operator response. Using the Indian Point Unit 2 Significance Determination Process Phase 2 "Transient with Power Conversion System Available" worksheet, this finding was determined to be of very low safety significance.
Inspection Report# : 2003013(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation TS 6.8.1 VIOLATION - FAILURE TO ADHERE TO EMERGENCY OPERATING PROCEDURE ES-0.1, CONTINUOUS ACTION STEP 1.0 ON AUGUST 3, 2003 The team identified a violation involving the failure of an operating crew to adhere to a continuous action step of Emergency Operating Procedure ES-0.1, "Reactor Trip Response," resulting in an avoidable plant transient. Specifically, in response to the reactor trip and partial loss of offsite power (LOOP) event on August 3, 2003, the Unit 2 operating crew did not correctly implement continuous action step 1 of ES-0.1, which led to the cycling of the pressurizer power-operated relief valves (PORVs) ten times, complicating reactor coolant system (RCS) pressure control.
This finding is greater than minor because it affected the Initiating Events Cornerstone and could reasonably be viewed as a precursor to a more significant event, in that, the failure to implement established procedures could place the reactor outside its design envelope and, for this particular event, the repeated cycling of the PORVs could have resulted in a loss of coolant event had a PORV stuck open. This finding is of very low safety significance because all mitigation systems were available during the event and was treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2003013(pdf)
Significance:        May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER EMERGENT WORK PACKAGE INSTRUCTIONS FOR 22 STEAM GENERATOR LEVEL BISTABLE REPLACEMENT On February 7, 2003, a self-revealing finding involved inadequate emergent work instructions that resulted in an electrical short during replacement of the 22 steam generator low level bistable. The electrical short caused a breaker trip on circuit 10 of instrument bus 21and the resultant loss of electrical power to the pressurizer level and reactor coolant system pressure control channels (failed low). The inadequate work instructions is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V, since the instructions did not account for consideration of performing this replacement with the circuit de-energized or the proximity to other reactor protection system relays.
The performance issue is more than minor since the operators were required to take action to restore reactor coolant system pressure and pressurizer level to preclude a reactor trip. The finding involves the initiating events cornerstone in that it increased the likelihood of upset in plant stability and it involves human error during the planning of an emergent work activity. This finding is considered to be of very low safety significance in that in accordance with NRC Manual Chapter 0609, Appendix A, the finding did not contribute to the likelihood of a secondary or primary LOCA initiator and it did not contribute to either a reactor trip or mitigation system unavailability.
Inspection Report# : 2003003(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 2 of 7 Mitigating Systems Significance:        Jan 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO CONDUCT SIMULATOR TESTING IN ACCORDANCE WITH ANSI/ANS 3.5-1985 The inspectors identified that simulator performance testing did not meet the standards as specified in ANSI/ANS 3.5-1985 in that: (1) "best estimate" data for the simulator testing was not used; (2) all required key parameters during the simulator test were not recorded; and (3) simulator differences identified during testing were not documented and justified.
This finding is more than minor because it affects the human performance (human error) attribute of the mitigating systems cornerstone. More specifically, improperly conducted simulator testing resulted, in part, in not identifying replication issues for steam generator pressure and cold leg temperature. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE SIMULATOR TO DEMONSTRATE EXPECTED PLANT RESPONSE TO TRANSIENT CONDITIONS The inspectors identified a non-cited violation of 10 CFR 55.46(c)(1), involving the failure of the simulator to correctly replicate key parameters such as steam generator pressure and cold leg temperature (Tcold) during a loss of all reactor coolant pumps. Additionally, the plant decay heat load was not correctly modeled which contributed to inappropriate operator actions during the August 3, 2003, plant trip.
This finding is more than minor because it affected the human performance (human error) attribute of the mitigating systems cornerstone. Not correctly replicating the plant's response on the simulator provides the potential for negative operator training. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation A Green NCV was identified for failure to take appropriate corrective actions for Gas Turbine 1 An NCV of 10 CFR 50.65 (a)(1) was identified when Entergy failed to take appropriate corrective actions when the #1 Gas Turbine (GT1) exceeded its maintenance rule (a)(1) reliability monitoring goal. This finding was greater than minor because it affected the reliability of GT1 which is used to mitigate the consequences of a station blackout. This issue was evaluated using the significance determination process and determined to be of very low significance (Green) since the redundant train was always available to perform the GT safety functions.
Inspection Report# : 2003012(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: FIN Finding THE PERFORMANCE FINDING INVOLVED INADEQUATE SHORT TERM CORRECTIVE ACTIONS ASSOCIATED WITH FIRE LEAKS ON A FIRE HEADER IN THE UNIT 1 TURBINE BUILDING The inspectors identified a finding involving inadequate corrective actions associated with multiple leaks on a six-inch fire header in the Unit 1 turbine building. On September 10, 2003, an 80 gallon per minute fire header leak occurred that operators isolated by depressurizing the entire fire water suppression system at Unit 2 for approximately three hours. This leak occurred approximately one foot from a similar through-wall leak which occurred on July 16, 2003.
This performance issue is considered more than minor based on example 4.f. in MC 0612 Appendix E. The performance finding involves the Mitigating Systems Cornerstone objective of fire suppression system availability to respond to fires. The finding is very low risk significance based upon the results from the fire protection risk significance screening methodology (FPRSSM). The finding impacts both manual suppression capability and automatic suppression capability.
Inspection Report# : 2003011(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 3 of 7 Significance:        Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B, CRITERION V. THE PROCEDURAL STEPS FOR THE INSTALLATION OF A FLEXIBLE COUPLING WERE NOT ADEQUATE TO VERIFY THAT THE COMPONENT WAS PROPERLY INSTALLED.
The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V. In November 2002, a maintenance work instruction to install a 21 emergency diesel generator (EDG) service water supply flexible coupling did not include critical installation steps per the vendor manual. This resulted in a significant service water leak from the expansion joint on August 14, 2003.
This finding is greater than minor since if left uncorrected, it could be a more significant safety concern as this type of flexible coupling is used on all three EDGs. The inspectors determined that the expansion joint leakage was of a very low safety significance since it did not adversely impact service water cooling to the emergency diesel generator or the overall service water system cooling capability, did not impact equipment and functions associated with internal flooding in the diesel generator room, and did not result in a loss of service water or emergency power safety function that contributed to internal flooding initiated events.
Inspection Report# : 2003011(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B. A DESIGN CHANGE PACKAGE DID NOT ACCURATELY REFLECT ACTUAL PLANT CONDITIONS AND RESULTED IN AN UNINTENDED PLANT TRANSIENT The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, involving the design change package (DCP-200105716-I) to replace a pressurizer level recorder which did not contain accurate design details. As a consequence, during installation of the design change an unintended plant transient challenged operators.
This finding is greater than minor based upon NRC Manual Chapter 0612, Appendix E, example 4.b. This finding is of very low safety significance. The finding did contribute to the likelihood of a reactor trip; however, it did not impact the availability of mitigation equipment, increase the likelihood of a primary or secondary system LOCA, or increase the likelihood of an internal fire or flood.
Inspection Report# : 2003011(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation EQUIPMENT TAGOUT TO RESTORE THE 22 SEAL INJECTION FILTER WERE INADEQUATE TO MAINTAIN PROPER CONFIGURATION CONTROL OF THE SYSTEM The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, involving an incomplete procedure for restoring to service the 22 seal injection filter from maintenance. The consequence was an approximate 70 gallon per minute chemical volume and control system leak through an open vent valve which lasted for approximately two minutes before operators identified and shut the vent valve.
This finding is more than minor since it adversely impacted the Mitigating System Cornerstone objective of safety system capability and availability with respect to the attributes of configuration control and procedural quality. The inadequate restoration procedure resulted in a significant chemical and volume control system leak (the capacity of one coolant charging pump) that degraded normal charging flow and emergency boration capability for a short period of time. The finding is of very low safety significance since it did not result in a loss of emergency boration safety function.
Inspection Report# : 2003011(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: FIN Finding INADEQUATE OPERABILITY EVALUATION FOR THE 13.8 KV SYSTEM The inspector identified that the licensee's operability evaluation during a 13.8 KV system reduced voltage test was inadequate. The operability evaluation did not evaluate accident load carrying capability as defined in the technical specification basis and it did not address communications and protocols between the distribution company and the licensee to restore from the test in a timely manner. NRC Manual Chapter 9900 states that when a system's capability is degraded to a point where it cannot perform with reasonable assurance of reliability, the system should be judged inoperable.
The finding was more than minor because it impacted the attribute of the mitigating system cornerstone objective. Specifically, the cornerstone objective is to ensure that the 13.8 KV system is capable of performing its safety function during a postulated loss of normal power event without undesirable consequences. This finding was determined to be of low safety significance because it did not result in the actual loss of 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 4 of 7 the offsite power supply safety function.
Inspection Report# : 2003007(pdf)
Significance:        Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions associated with an unauthorized modification to the No. 22 component cooling water pump.
The inspector identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI. The licensee did not evaluate and take effective corrective actions associated with a material substitution for the 22 component cooling water pump inboard bearing oil level indication system.
The bearing oil level indication system contributed to the failure of the #22 CCW pump on December 5, 2002.
This finding is greater than minor since it is associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective. The inspectors conducted a Phase 1 SDP screening and determined that the failure to take effective corrective action on
#22 CCW pump was of a very low safety significance since the redundant train components were operable and unaffected by this unauthorized modification.
Inspection Report# : 2003007(pdf)
Significance:        May 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS ASSOCIATED WITH THE 23 EDG LOAD SWINGS BETWEEN MAY 2000 AND FEBRUARY 2003 The inspectors identified that ineffective corrective actions resulted in repetitive surveillance test failures of the 23 emergency diesel generator between December 2001 and February 2003. This finding is considered a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI. The finding is more than minor because the surveillance test failures impacted the availability of one train of emergency AC power source. This finding was of very low risk significance because the repetitive failures did not result in an actual loss of function for the emergency AC power.
Inspection Report# : 2003003(pdf)
Significance:        May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation POST-WORK TEST INADEQUAATE FOR 22 BORIC ACID TRANSFER PUMP BORIC ACID FILTER STOP VALVE A self-revealing event was identified on February 26, 2003, when operators observed no boric acid flow to the reactor vessel via the No. 22 boric acid transfer pump (BATP). It was determined that during preventative maintenance activities in March 2001, the post-work test on the No. 22 BATP outlet valve to the boric acid filter stop was inadequate to identify that the valve finger plate was installed upside down. This finding is considered a non-cited violation of 10 CFR 50 Appendix B, Criterion V. This event is considered more than minor because the improperly installed valve plate affected the availability of one train of emergency boration. This is considered to be of very low risk significance in accordance with NRC MC 0609 Appendix A, since the emergency boration function was not lost due to this performance issue.
Inspection Report# : 2003003(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: VIO Violation VIOLATION OF THE APPROVED FIRE PROTECTION PROGRAM/THREE-HOUR RATED WALL CONSTRUCTED TO SEPARATE THE CONTROL BUILDING FROM THE TURBINE BUILDING WHITE - The team identified a violation of License Condition 2.K of Facility Operating License DPR-26. License Condition 2.K requires that Entergy implement and maintain in effect all provisions of the NRC approved fire protection program, which states that a three-hour rated wall will be constructed to separate the control building from the turbine building. In 1978, to meet the three-hour rating, the wall was to have been built in accordance with the design specification Underwriters Laboratories (UL) U902. Contrary to the above, in February 2002, the wall was found not to be constructed in accordance with UL U902.
The combined effect of the identified deficiencies was that, as of February 2002, passages existed through both the outer brick and inner portions of the wall. If a significant amount of smoke and gasses were to penetrate the wall, this could result in the CCR becoming uninhabitable, causing the operators to resort to using the Alternate Safe Shutdown System. These conditions did not represent a three-hour fire barrier. The NRC risk assessment, using Phase 2 of the NRC Fire SDP described in MC 0609, Appendix F, considered the wall a moderately degraded fire barrier having low to moderate safety significance (White). Until repairs could be completed, Entergy established a compensatory fire watch in accordance with the IP2 fire protection program.
Entergy actions in identifying original construction deficiencies in the CCR west inner wall in February 2002 were commendable. However, 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 5 of 7 the corrective actions taken were not fully effective in restoring the wall to its three-hour rated design configuration. Additionally, the initial extent of condition was not sufficient to identify other degraded fire barrier walls.
[Final Significance Determination and Notice of Violation docketed in NRC letter, dated November 8, 2002. Entergy response to NOV dated December 9, 2002]
Inspection Report# : 2002010(pdf)
Inspection Report# : 2004003(pdf)
Barrier Integrity Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions for repetitive failures of the plant vent noble gas effluent monitor The team identified a finding of very low safety significance (Green) regarding the licensee's inadequate corrective actions for repetitive failures of a TS required surveillance of the plant vent noble gas effluent monitor. Since July 2002, the monitor has failed the surveillance test five of the six times performed. The performance deficiency associated with this findings was inadequate corrective actions for repetitive failures of a TS required surveillance. This finding is more than minor significance because the R-27 radiation monitor was removed from service for periods in excess of twenty-four hours as a result of inadequate corrective actions. The finding was evaluated using the EP SDP, and was screened to be of very low safety significance, because there were alternate monitoring methods available in the event of an accident.
Inspection Report# : 2004003(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to implement appropriate corrective actions for degraded Technical Support Center Batteries The inspectors identified a findings of very low safety significance (Green) regarding the licensee's failure to implement appropriate corrective actions for degraded TSC batteries. The perofrmance deficiency associated with this findings was the failure to take timely and effective corrective actions for the degraded TSC batteries. The degraded batteries adversely impacted the Non-Risk Significant Planning Standard, as described in 10 CFR 50.47(b)(8), which requires adequate facilities and equipment be maintained to support emergency response. This finding was more than minor significance because the batteries were allowed to remain in a degraded stated in excess of twenty-four hours without adequate measures to ensure that their TSC support function would be maintained. The finding is of very low safety significance, because the subsequent analysis indicated that the battery banks remained functional in this condition.
Inspection Report# : 2004003(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE EOF UPSs ON AUGUST 14 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the August 14, 2003, loss of off-site power event which revealed that Entergy did not have a preventive maintenance program in place to ensure the continued functionality of the numerous un-interruptible power supplies in the Emergency Operations Facility (EOF) which provide back-up power to emergency response equipment.
This finding is considered greater than minor because a significant amount of the Unit 2 and Unit 3 emergency response organization communications equipment was non-functional on August 14 until off-site power was restored. However, this finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions from the EOF, TSC/OSC, and Unit 2 and 3 central control rooms, using back-up telephone communications.
Inspection Report# : 2003013(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 6 of 7 Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE UNIT 2 TSC DIESEL ON AUGUST 14, 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the failure of the Unit 2 TSC back-up diesel generator to function on August 14, 2003. The conditions which caused the diesel generator to fail to function involved electrical loading of the diesel generator in excess of its design capacity. This condition was initially identified in February 2000 and not resolved in a timely manner.
This finding is considered more than minor because a significant amount of TSC/OSC emergency response equipment, necessary to implement the Emergency Plan, was either de-energized by the Entergy staff because of the loss of sufficient air conditioning to ensure emergency response equipment would not be damaged due to overheating, or was without AC power because the diesel was non-functional. This finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions.
Inspection Report# : 2003013(pdf)
Occupational Radiation Safety Public Radiation Safety Significance:        May 15, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH PACKAGING PROCEDURES A self-revealing non-cited violation of 10 CFR 71.12 was identified for failure to comply with shipping cask package procedures. On February 6, 2003, a CNS 8-120 B cask was received from the Indian Point Energy Center at a consolidation facility in South Carolina with a bolt missing on the primary lid's pressure test port in violation of the cask use and maintenance procedures. This finding was more than minor in that it was associated with the Public Radiation Safety Cornerstone's attribute of procedures for transportation packages. The finding affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials contained in an NRC-approved Type B package released into the public domain. The finding was determined to be of very low safety significance in that the finding did not involve exceeding transportation radiation limits, there was no breach of the package during transit, and the issue was a Certificate of Compliance maintenance/use performance deficiency.
Inspection Report# : 2003003(pdf)
Physical Protection Miscellaneous Significance: N/A Dec 12, 2003 Identified By: NRC Item Type: FIN Finding The licensee's Corrective Action Program, used for identifying, tracking, prioritizing, and resolving deficiencies, was appropriately implemented in most cases.
The inspection team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program, evaluating and prioritizing issues, and implementing appropriate corrective actions. However, the inspectors identified some minor equipment problems that had not been identified in the corrective action program. The evaluation of problems wa generally adequate, but the inspectors identified two Green findings related to the failure to implement effective corrective acitons for degraded emergency preparedness equipment. These findings were determined not to involve violations of NRC requirements, however, they represented additional examples of the substantive cross-cutting issue in the problem identification and resolution area that was identified during the previous assessments. Based on interviews conducted during the inspection, station personnel felt free to identify safety issues and enter them into the corrective action 07/14/2004
 
1Q/2004 Inspection Findings - Indian Point 2                                                                                        Page 7 of 7 program.
Inspection Report# : 2004003(pdf)
Significance: TBD Apr 01, 2000 Identified By: Licensee Item Type: FIN Finding Contamination in Storm Drains Con Edison staff appropriately responded to the discovery of trace amounts of contamination in the Unit 1 storm drains and took proper actions to resolve the condition and to investigate the cause. The material was not associated with the Unit 2 steam generator event or any recent plant activities, and there was no radiological dose consequence due to the contamination.
Inspection Report# : 2000003(pdf)
Inspection Report# : 2001010(pdf)
Significance: TBD Apr 01, 2000 Identified By: NRC Item Type: FIN Finding Steam Generator Tube Leak Root Cause Evaluation Con Edison completed the investigation of the plant response to the February 15, 2000 steam generator tube leak. Corrective actions to address the causes of weaknesses in the plant response to the event were in progress at the end of the inspection period and NRC review will be the subject of an AIT follow-up team inspection. The results of the root cause investigation for the steam generator tube failure were not reviewed and are being provided by Con Edison to the NRC Office of Nuclear Reactor Regulation for review.
Inspection Report# : 2000003(pdf)
Last modified : May 05, 2004 07/14/2004
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                      Page 1 of 7 Indian Point 2 2Q/2004 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 5.4.1.d FOR THE FAILURE TO PERFORM A TRANSIENT COMBUSTIBLE EVALUATION FOR 330 GALLONS OF OIL TEMPORARILY STORED IN FIRE ZONE 6A The inspector identified a non-cited violation of Technical Specification (TS) 5.4.1.d. that requires, in part, that written procedures shall be implemented for the Fire Protection Program. The inspector determined that no transient combustible evaluation (TCE) was completed for approximately 330 gallons of lubricating oil stored in fire zone 6A, "Waste Drumming and Storage Station," contrary to Procedure ENN-DC-161, "Transient Combustible Program," step 5.2.3.
This finding is greater than minor because it represented a condition similar to example 4.k in Appendix E, IMC 0612, in that the as-found condition involved transient combustible material loading in excess of the Fire Hazard Analysis limit. The finding is of very low safety significance because it did not increase the likelihood of a fire, no credible fire scenario was identified due to the type of storage containers used, there were no intervening combustibles, and no credible fire ignition source was present.
Inspection Report# : 2004002(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE APPROPRIATE AND TIMELY CORRECTIVE ACTIONS TO ADDRESS THE REPEATED GRID-RELATED REACTOR TRIPS OF UNIT 2 This team-identified finding involves inadequate corrective actions for repeat Unit 2 reactor scrams attributed to grid-related faults and associated protective relaying failures. The lack of thorough evaluations and corrective actions on the part of Entergy, in cooperation with the responsible Transmission and Distribution Operator for the local area electrical grid, have resulted in an increased frequency of plant transients and consequential challenges to Unit 2 safety related systems and licensed operators.
This finding is greater than minor because it affects the Initiating Events Cornerstone and represents an increased likelihood of an event that challenges critical safety functions and operator response. Using the Indian Point Unit 2 Significance Determination Process Phase 2 "Transient with Power Conversion System Available" worksheet, this finding was determined to be of very low safety significance.
Inspection Report# : 2003013(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation TS 6.8.1 VIOLATION - FAILURE TO ADHERE TO EMERGENCY OPERATING PROCEDURE ES-0.1, CONTINUOUS ACTION STEP 1.0 ON AUGUST 3, 2003 The team identified a violation involving the failure of an operating crew to adhere to a continuous action step of Emergency Operating Procedure ES-0.1, "Reactor Trip Response," resulting in an avoidable plant transient. Specifically, in response to the reactor trip and partial loss of offsite power (LOOP) event on August 3, 2003, the Unit 2 operating crew did not correctly implement continuous action step 1 of ES-0.1, which led to the cycling of the pressurizer power-operated relief valves (PORVs) ten times, complicating reactor coolant system (RCS) pressure control.
This finding is greater than minor because it affected the Initiating Events Cornerstone and could reasonably be viewed as a precursor to a more significant event, in that, the failure to implement established procedures could place the reactor outside its design envelope and, for this particular event, the repeated cycling of the PORVs could have resulted in a loss of coolant event had a PORV stuck open. This finding is of very low safety significance because all mitigation systems were available during the event and was treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2003013(pdf)
Mitigating Systems
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                  Page 2 of 7 Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a condition adverse to quality which could impact EDG reliability.
The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality concerning emergency diesel generator (EDG) heat exchanger (HX) fouling.
This finding was more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring equipment availability and reliability of the EDG HXs to perform their intended safety function. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone. However, this finding was determined to have very low safety significance (Green) using the SDP Phase 1 screening worksheet for mitigating systems because the EDG HXs remained operable and capable of performing their intended safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: FIN Finding Failure to implement adequate corrective actions for low voltage conditions on the 13.8 kV system.
The inspectors identified a finding due to ineffective and untimely corrective actions associated with the 13.8 KV system during reduced voltage conditions.
This finding was determined to be greater than minor since it impacts the mitigating systems cornerstone objective of ensuring system reliability and capability. This finding was associated with the procedure quality attribute of that cornerstone. This finding was of very low safety significance since there was no loss of the normal offsite power supplies and the 13.8 KV system was not providing power to any safety-related loads during the degraded condition.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls during modifications to the recirculation sump.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for Entergy's failure to translate the emergency core cooling system (ECCS) design basis into recirculation sump modification instructions. Specifically, Entergy added penetration cover plates and alignment collars around several small pipes that penetrated the sump deck plating, and the annular gap between the collars and pipes exceeded the sump screen size.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (loss-of-coolant accidents) (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a recirculation sump screen bypass path.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality. Specifically, Entergy did not promptly identify and correct a recirculation sump bypass path and containment debris that had the potential to adversely impact ECCS during containment recirculation.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a Technical Specification Surveillance Requirement The inspector identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.3.1.1. that requires, in part, that a channel check be performed every 12 hours on the feedwater flow instrumentation in the central control room. This requirement had not been met since the licensee implemented the Improved Technical Specifications in December of 2003.
This finding is greater than minor because it represented a condition similar to example 1.c in Appendix E, IMC 0612, in that the surveillance was not
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                      Page 3 of 7 performed per Technical Specifications from December 12, 2003 through June 8, 2004. The finding is of very low safety significance because the feedwater flow instruments met the surveillance criteria when subsequently performed, and did not render the mitigating equipment inoperable.
Inspection Report# : 2004006(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE FAILURE TO TAKE APPROPRIATE CORRECTIVE ACTIONS TO ENSURE THE RELIABILITY AND AVAILABILITY OF GAS TURBINE NO. 1 A finding was identified involving untimely corrective actions which contributed to increased unavailability of Gas Turbine 1 (GT-1) which is considered a mitigating system. Specifically, GT-1 was not available for approximately 116 hours due to the failure and subsequent replacement of the starting diesel battery charger.
This finding was determined to be greater than minor since it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring system reliability and availability of systems that are used to prevent undesirable consequences due to initiating events. GT-1 is credited as an alternate AC power source for both station blackout and Appendix R fire scenarios. This finding was considered of very low safety significance because there was no actual loss of safety function for this mitigating system, since GT-3 was available while GT-1 was inoperable. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE IMPROPER CONTROL OF AN OUT-GOING 345 KV FEEDER BREAKER DURING A FEEDER OUTAGE The inspector identified that the control room operators placed the 345KV ring bus in a configuration that would challenge the availability of mitigating systems in the event of an off-site electrical transient. Specifically, in the event that a 345KV feeder fault caused a loss-of-load plant trip, two of the four 480 volt safety buses would require operator action to restore.
This finding is greater than minor since it is associated with the configuration control attribute of the mitigating systems cornerstone and that it impacted the cornerstone objective of ensuring the availability of mitigating systems. With the ring bus aligned with both output breakers shut and one feeder out of service, a subsequent fault on the remaining feeder would have resulted in a plant trip with the fast transfer blocked. This would de-energize vital busses 2A and 3A causing a loss of power to one of two motor-driven auxiliary feed pumps, one of three safety injection pumps, two of five containment fan cooler units, one of two residual heat removal pumps and two of six service water pumps. Manual operator action would then be required to restore this equipment. This finding is considered of very low safety significance because it did not result in an actual loss of safety function of any mitigating systems. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50, APPENDIX B, CRITERION III "DESIGN CONTROL" FOR THE FAILURE TO IMPLEMENT APPROPRIATE DESIGN CONTROLS FOR A MODIFICATION MADE TO THE OTDT CONTROLLER The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control." A component was modified during the replacement of a safety-related controller in the over-temperature delta-temperature (OTDT) circuitry of the reactor protection system without a formal modification package.
This finding was determined to be greater than minor since it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring reactor protection system reliability. Specifically, the failure to use a derated resistor to modify the circuit card had an adverse impact on the reliability of the OTDT controller. This finding is considered of very low safety significance since it did not result in the actual loss of safety function of a system. This issue did not impact fire, flooding, seismic, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO CONDUCT SIMULATOR TESTING IN ACCORDANCE WITH ANSI/ANS 3.5-1985 The inspectors identified that simulator performance testing did not meet the standards as specified in ANSI/ANS 3.5-1985 in that: (1) "best estimate" data for the simulator testing was not used; (2) all required key parameters during the simulator test were not recorded; and (3) simulator differences identified during testing were not documented and justified.
This finding is more than minor because it affects the human performance (human error) attribute of the mitigating systems cornerstone. More
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                    Page 4 of 7 specifically, improperly conducted simulator testing resulted, in part, in not identifying replication issues for steam generator pressure and cold leg temperature. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE SIMULATOR TO DEMONSTRATE EXPECTED PLANT RESPONSE TO TRANSIENT CONDITIONS The inspectors identified a non-cited violation of 10 CFR 55.46(c)(1), involving the failure of the simulator to correctly replicate key parameters such as steam generator pressure and cold leg temperature (Tcold) during a loss of all reactor coolant pumps. Additionally, the plant decay heat load was not correctly modeled which contributed to inappropriate operator actions during the August 3, 2003, plant trip.
This finding is more than minor because it affected the human performance (human error) attribute of the mitigating systems cornerstone. Not correctly replicating the plant's response on the simulator provides the potential for negative operator training. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation A Green Non-cited violation was identified for failure to take appropriate corrective actions for Gas Turbine 1 An NCV of 10 CFR 50.65 (a)(1) was identified when Entergy failed to take appropriate corrective actions when the #1 Gas Turbine (GT1) exceeded its maintenance rule (a)(1) reliability monitoring goal. This finding was greater than minor because it affected the reliability of GT1 which is used to mitigate the consequences of a station blackout. This issue was evaluated using the significance determination process and determined to be of very low significance (Green) since the redundant train was always available to perform the GT safety functions.
Inspection Report# : 2003012(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B. A DESIGN CHANGE PACKAGE DID NOT ACCURATELY REFLECT ACTUAL PLANT CONDITIONS AND RESULTED IN AN UNINTENDED PLANT TRANSIENT The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, involving the design change package (DCP-200105716-I) to replace a pressurizer level recorder which did not contain accurate design details. As a consequence, during installation of the design change an unintended plant transient challenged operators.
This finding is greater than minor based upon NRC Manual Chapter 0612, Appendix E, example 4.b. This finding is of very low safety significance. The finding did contribute to the likelihood of a reactor trip; however, it did not impact the availability of mitigation equipment, increase the likelihood of a primary or secondary system LOCA, or increase the likelihood of an internal fire or flood.
Inspection Report# : 2003011(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: FIN Finding THE PERFORMANCE FINDING INVOLVED INADEQUATE SHORT TERM CORRECTIVE ACTIONS ASSOCIATED WITH FIRE LEAKS ON A FIRE HEADER IN THE UNIT 1 TURBINE BUILDING The inspectors identified a finding involving inadequate corrective actions associated with multiple leaks on a six-inch fire header in the Unit 1 turbine building. On September 10, 2003, an 80 gallon per minute fire header leak occurred that operators isolated by depressurizing the entire fire water suppression system at Unit 2 for approximately three hours. This leak occurred approximately one foot from a similar through-wall leak which occurred on July 16, 2003.
This performance issue is considered more than minor based on example 4.f. in MC 0612 Appendix E. The performance finding involves the Mitigating Systems Cornerstone objective of fire suppression system availability to respond to fires. The finding is very low risk significance based upon the results from the fire protection risk significance screening methodology (FPRSSM). The finding impacts both manual suppression capability and automatic suppression capability.
Inspection Report# : 2003011(pdf)
Significance:        Sep 27, 2003
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                      Page 5 of 7 Identified By: NRC Item Type: NCV NonCited Violation NCV OF 10 CFR 50, APP B, CRITERION V. THE PROCEDURAL STEPS FOR THE INSTALLATION OF A FLEXIBLE COUPLING WERE NOT ADEQUATE TO VERIFY THAT THE COMPONENT WAS PROPERLY INSTALLED.
The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V. In November 2002, a maintenance work instruction to install a 21 emergency diesel generator (EDG) service water supply flexible coupling did not include critical installation steps per the vendor manual. This resulted in a significant service water leak from the expansion joint on August 14, 2003.
This finding is greater than minor since if left uncorrected, it could be a more significant safety concern as this type of flexible coupling is used on all three EDGs. The inspectors determined that the expansion joint leakage was of a very low safety significance since it did not adversely impact service water cooling to the emergency diesel generator or the overall service water system cooling capability, did not impact equipment and functions associated with internal flooding in the diesel generator room, and did not result in a loss of service water or emergency power safety function that contributed to internal flooding initiated events.
Inspection Report# : 2003011(pdf)
Significance:        Sep 27, 2003 Identified By: NRC Item Type: NCV NonCited Violation EQUIPMENT TAGOUT TO RESTORE THE 22 SEAL INJECTION FILTER WERE INADEQUATE TO MAINTAIN PROPER CONFIGURATION CONTROL OF THE SYSTEM The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, involving an incomplete procedure for restoring to service the 22 seal injection filter from maintenance. The consequence was an approximate 70 gallon per minute chemical volume and control system leak through an open vent valve which lasted for approximately two minutes before operators identified and shut the vent valve.
This finding is more than minor since it adversely impacted the Mitigating System Cornerstone objective of safety system capability and availability with respect to the attributes of configuration control and procedural quality. The inadequate restoration procedure resulted in a significant chemical and volume control system leak (the capacity of one coolant charging pump) that degraded normal charging flow and emergency boration capability for a short period of time. The finding is of very low safety significance since it did not result in a loss of emergency boration safety function.
Inspection Report# : 2003011(pdf)
Barrier Integrity Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to evaluate the degraded condition of the Technical Support Center batteries The inspectors identified a findings of very low safety significance (Green) regarding the licensee's failure to implement appropriate corrective actions for degraded TSC batteries. The perofrmance deficiency associated with this findings was the failure to take timely and effective corrective actions for the degraded TSC batteries. The degraded batteries adversely impacted the Non-Risk Significant Planning Standard, as described in 10 CFR 50.47(b)
(8), which requires adequate facilities and equipment be maintained to support emergency response. This finding was more than minor significance because the batteries were allowed to remain in a degraded stated in excess of twenty-four hours without adequate measures to ensure that their TSC support function would be maintained. The finding is of very low safety significance, because the subsequent analysis indicated that the battery banks remained functional in this condition.
Inspection Report# : 2004003(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and address causes fo repetitive surveillance test failures of the plant vent noble gas effluent monitor.
The team identified a finding of very low safety significance (Green) regarding the licensee's inadequate corrective actions for repetitive failures of a TS required surveillance of the plant vent noble gas effluent monitor. Since July 2002, the monitor has failed the surveillance test five of the six times performed. The performance deficiency associated with this findings was inadequate corrective actions for repetitive failures of a TS required surveillance. This finding is more than minor significance because the R-27 radiation monitor was removed from service for periods in excess of twenty-four hours as a result of inadequate corrective actions. The finding was evaluated using the EP SDP, and was screened to be of very low safety significance, because there were alternate monitoring methods available in the event of an accident.
Inspection Report# : 2004003(pdf)
 
2Q/2004 Inspection Findings - Indian Point 2                                                                                                    Page 6 of 7 Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE EOF UPSs ON AUGUST 14 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the August 14, 2003, loss of off-site power event which revealed that Entergy did not have a preventive maintenance program in place to ensure the continued functionality of the numerous un-interruptible power supplies in the Emergency Operations Facility (EOF) which provide back-up power to emergency response equipment.
This finding is considered greater than minor because a significant amount of the Unit 2 and Unit 3 emergency response organization communications equipment was non-functional on August 14 until off-site power was restored. However, this finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions from the EOF, TSC/OSC, and Unit 2 and 3 central control rooms, using back-up telephone communications.
Inspection Report# : 2003013(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE UNIT 2 TSC DIESEL ON AUGUST 14, 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the failure of the Unit 2 TSC back-up diesel generator to function on August 14, 2003. The conditions which caused the diesel generator to fail to function involved electrical loading of the diesel generator in excess of its design capacity. This condition was initially identified in February 2000 and not resolved in a timely manner.
This finding is considered more than minor because a significant amount of TSC/OSC emergency response equipment, necessary to implement the Emergency Plan, was either de-energized by the Entergy staff because of the loss of sufficient air conditioning to ensure emergency response equipment would not be damaged due to overheating, or was without AC power because the diesel was non-functional. This finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions.
Inspection Report# : 2003013(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Dec 12, 2003 Identified By: NRC Item Type: FIN Finding The licensee's Corrective Action Program, used for identifying, tracking, prioritizing, and resolving deficiencies, was appropriately implemented in most cases.
Problem Identification & Resolution Team Inspection Summary: The inspection team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program, evaluating and prioritizing issues, and implementing appropriate corrective actions.
However, the inspectors identified some minor equipment problems that had not been identified in the corrective action program. The evaluation of problems was generally adequate, but the inspectors identified two Green findings related to the failure to implement effective corrective acitons for degraded emergency preparedness equipment. These findings were determined not to involve violations of NRC requirements, however, they represented additional examples of the substantive cross-cutting issue in the problem identification and resolution area that was identified during the previous assessments. Based on interviews conducted during the inspection, station personnel felt free to identify safety issues and enter them into the
 
2Q/2004 Inspection Findings - Indian Point 2 Page 7 of 7 corrective action program.
Inspection Report# : 2004003(pdf)
Last modified : September 08, 2004
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 1 of 6 Indian Point 2 3Q/2004 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 5.4.1.d FOR THE FAILURE TO PERFORM A TRANSIENT COMBUSTIBLE EVALUATION FOR 330 GALLONS OF OIL TEMPORARILY STORED IN FIRE ZONE 6A The inspector identified a non-cited violation of Technical Specification (TS) 5.4.1.d. that requires, in part, that written procedures shall be implemented for the Fire Protection Program. The inspector determined that no transient combustible evaluation (TCE) was completed for approximately 330 gallons of lubricating oil stored in fire zone 6A, "Waste Drumming and Storage Station," contrary to Procedure ENN-DC-161, "Transient Combustible Program," step 5.2.3.
This finding is greater than minor because it represented a condition similar to example 4.k in Appendix E, IMC 0612, in that the as-found condition involved transient combustible material loading in excess of the Fire Hazard Analysis limit. The finding is of very low safety significance because it did not increase the likelihood of a fire, no credible fire scenario was identified due to the type of storage containers used, there were no intervening combustibles, and no credible fire ignition source was present.
Inspection Report# : 2004002(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE APPROPRIATE AND TIMELY CORRECTIVE ACTIONS TO ADDRESS THE REPEATED GRID-RELATED REACTOR TRIPS OF UNIT 2 This team-identified finding involves inadequate corrective actions for repeat Unit 2 reactor scrams attributed to grid-related faults and associated protective relaying failures. The lack of thorough evaluations and corrective actions on the part of Entergy, in cooperation with the responsible Transmission and Distribution Operator for the local area electrical grid, have resulted in an increased frequency of plant transients and consequential challenges to Unit 2 safety related systems and licensed operators.
This finding is greater than minor because it affects the Initiating Events Cornerstone and represents an increased likelihood of an event that challenges critical safety functions and operator response. Using the Indian Point Unit 2 Significance Determination Process Phase 2 "Transient with Power Conversion System Available" worksheet, this finding was determined to be of very low safety significance.
Inspection Report# : 2003013(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation TS 6.8.1 VIOLATION - FAILURE TO ADHERE TO EMERGENCY OPERATING PROCEDURE ES-0.1, CONTINUOUS ACTION STEP 1.0 ON AUGUST 3, 2003 The team identified a violation involving the failure of an operating crew to adhere to a continuous action step of Emergency Operating Procedure ES-0.1, "Reactor Trip Response," resulting in an avoidable plant transient. Specifically, in response to the reactor trip and partial loss of offsite power (LOOP) event on August 3, 2003, the Unit 2 operating crew did not correctly implement continuous action step 1 of ES-0.1, which led to the cycling of the pressurizer power-operated relief valves (PORVs) ten times, complicating reactor coolant system (RCS) pressure control.
This finding is greater than minor because it affected the Initiating Events Cornerstone and could reasonably be viewed as a precursor to a more significant event, in that, the failure to implement established procedures could place the reactor outside its design envelope and, for this particular event, the repeated cycling of the PORVs could have resulted in a loss of coolant event had a PORV stuck open. This finding is of very low safety significance because all mitigation systems were available during the event and was treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2003013(pdf)
Mitigating Systems
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 2 of 6 Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate and timely corrective actions for known deficiencies in the control program(s) and installation of safety related electrical cables and raceways.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address conditions adverse to quality concerning one example of resolution of Data Verification Transfer Report (DVTR) Items/Operability Assessments; and two examples of configuration control of electrical raceways and cables.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls for electrical cable and raceway installations.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for Entergy's failure to implement appropriate design controls for the installation of safety related electrical cables and raceways.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly control cable separation program documents.
Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for Entergy's failure to properly control the cable separation program documents. These documents include some reports never being reviewed, approved, and signed off as well as documents used in part for design specifications and DBD work not entered into the document control program to ensure retrieveability.
Inspection Report# : 2004009(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: FIN Finding Failure to implement adequate corrective actions for low voltage conditions on the 13.8 kV system.
The inspectors identified a finding due to ineffective and untimely corrective actions associated with the 13.8 KV system during reduced voltage conditions.
This finding was determined to be greater than minor since it impacts the mitigating systems cornerstone objective of ensuring system reliability and capability. This finding was associated with the procedure quality attribute of that cornerstone. This finding was of very low safety significance since there was no loss of the normal offsite power supplies and the 13.8 KV system was not providing power to any safety-related loads during the degraded condition.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls during modifications to the recirculation sump.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for Entergy's failure to translate the emergency core cooling system (ECCS) design basis into recirculation sump modification instructions. Specifically, Entergy added penetration cover plates and alignment collars around several small pipes that penetrated the sump deck plating, and the annular gap between the collars and pipes exceeded the sump screen size.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (loss-of-coolant accidents) (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                        Page 3 of 6 Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a recirculation sump screen bypass path.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality. Specifically, Entergy did not promptly identify and correct a recirculation sump bypass path and containment debris that had the potential to adversely impact ECCS during containment recirculation.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a condition adverse to quality which could impact EDG reliability.
The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality concerning emergency diesel generator (EDG) heat exchanger (HX) fouling.
This finding was more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring equipment availability and reliability of the EDG HXs to perform their intended safety function. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone. However, this finding was determined to have very low safety significance (Green) using the SDP Phase 1 screening worksheet for mitigating systems because the EDG HXs remained operable and capable of performing their intended safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a Technical Specification Surveillance Requirement The inspector identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.3.1.1. that requires, in part, that a channel check be performed every 12 hours on the feedwater flow instrumentation in the central control room. This requirement had not been met since the licensee implemented the Improved Technical Specifications in December of 2003.
This finding is greater than minor because it represented a condition similar to example 1.c in Appendix E, IMC 0612, in that the surveillance was not performed per Technical Specifications from December 12, 2003 through June 8, 2004. The finding is of very low safety significance because the feedwater flow instruments met the surveillance criteria when subsequently performed, and did not render the mitigating equipment inoperable.
Inspection Report# : 2004006(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE FAILURE TO TAKE APPROPRIATE CORRECTIVE ACTIONS TO ENSURE THE RELIABILITY AND AVAILABILITY OF GAS TURBINE NO. 1 A finding was identified involving untimely corrective actions which contributed to increased unavailability of Gas Turbine 1 (GT-1) which is considered a mitigating system. Specifically, GT-1 was not available for approximately 116 hours due to the failure and subsequent replacement of the starting diesel battery charger.
This finding was determined to be greater than minor since it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring system reliability and availability of systems that are used to prevent undesirable consequences due to initiating events. GT-1 is credited as an alternate AC power source for both station blackout and Appendix R fire scenarios. This finding was considered of very low safety significance because there was no actual loss of safety function for this mitigating system, since GT-3 was available while GT-1 was inoperable. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 4 of 6 Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE IMPROPER CONTROL OF AN OUT-GOING 345 KV FEEDER BREAKER DURING A FEEDER OUTAGE The inspector identified that the control room operators placed the 345KV ring bus in a configuration that would challenge the availability of mitigating systems in the event of an off-site electrical transient. Specifically, in the event that a 345KV feeder fault caused a loss-of-load plant trip, two of the four 480 volt safety buses would require operator action to restore.
This finding is greater than minor since it is associated with the configuration control attribute of the mitigating systems cornerstone and that it impacted the cornerstone objective of ensuring the availability of mitigating systems. With the ring bus aligned with both output breakers shut and one feeder out of service, a subsequent fault on the remaining feeder would have resulted in a plant trip with the fast transfer blocked. This would de-energize vital busses 2A and 3A causing a loss of power to one of two motor-driven auxiliary feed pumps, one of three safety injection pumps, two of five containment fan cooler units, one of two residual heat removal pumps and two of six service water pumps. Manual operator action would then be required to restore this equipment. This finding is considered of very low safety significance because it did not result in an actual loss of safety function of any mitigating systems. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50, APPENDIX B, CRITERION III "DESIGN CONTROL" FOR THE FAILURE TO IMPLEMENT APPROPRIATE DESIGN CONTROLS FOR A MODIFICATION MADE TO THE OTDT CONTROLLER The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control." A component was modified during the replacement of a safety-related controller in the over-temperature delta-temperature (OTDT) circuitry of the reactor protection system without a formal modification package.
This finding was determined to be greater than minor since it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring reactor protection system reliability. Specifically, the failure to use a derated resistor to modify the circuit card had an adverse impact on the reliability of the OTDT controller. This finding is considered of very low safety significance since it did not result in the actual loss of safety function of a system. This issue did not impact fire, flooding, seismic, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO CONDUCT SIMULATOR TESTING IN ACCORDANCE WITH ANSI/ANS 3.5-1985 The inspectors identified that simulator performance testing did not meet the standards as specified in ANSI/ANS 3.5-1985 in that: (1) "best estimate" data for the simulator testing was not used; (2) all required key parameters during the simulator test were not recorded; and (3) simulator differences identified during testing were not documented and justified.
This finding is more than minor because it affects the human performance (human error) attribute of the mitigating systems cornerstone. More specifically, improperly conducted simulator testing resulted, in part, in not identifying replication issues for steam generator pressure and cold leg temperature. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE SIMULATOR TO DEMONSTRATE EXPECTED PLANT RESPONSE TO TRANSIENT CONDITIONS The inspectors identified a non-cited violation of 10 CFR 55.46(c)(1), involving the failure of the simulator to correctly replicate key parameters such as steam generator pressure and cold leg temperature (Tcold) during a loss of all reactor coolant pumps. Additionally, the plant decay heat load was not correctly modeled which contributed to inappropriate operator actions during the August 3, 2003, plant trip.
This finding is more than minor because it affected the human performance (human error) attribute of the mitigating systems cornerstone. Not correctly replicating the plant's response on the simulator provides the potential for negative operator training. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                        Page 5 of 6 Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation A Green Non-cited violation was identified for failure to take appropriate corrective actions for Gas Turbine 1 An NCV of 10 CFR 50.65 (a)(1) was identified when Entergy failed to take appropriate corrective actions when the #1 Gas Turbine (GT1) exceeded its maintenance rule (a)(1) reliability monitoring goal. This finding was greater than minor because it affected the reliability of GT1 which is used to mitigate the consequences of a station blackout. This issue was evaluated using the significance determination process and determined to be of very low significance (Green) since the redundant train was always available to perform the GT safety functions.
Inspection Report# : 2003012(pdf)
Barrier Integrity Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to evaluate the degraded condition of the Technical Support Center batteries The inspectors identified a findings of very low safety significance (Green) regarding the licensee's failure to implement appropriate corrective actions for degraded TSC batteries. The perofrmance deficiency associated with this findings was the failure to take timely and effective corrective actions for the degraded TSC batteries. The degraded batteries adversely impacted the Non-Risk Significant Planning Standard, as described in 10 CFR 50.47(b)(8), which requires adequate facilities and equipment be maintained to support emergency response. This finding was more than minor significance because the batteries were allowed to remain in a degraded stated in excess of twenty-four hours without adequate measures to ensure that their TSC support function would be maintained. The finding is of very low safety significance, because the subsequent analysis indicated that the battery banks remained functional in this condition.
Inspection Report# : 2004003(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and address causes fo repetitive surveillance test failures of the plant vent noble gas effluent monitor.
The team identified a finding of very low safety significance (Green) regarding the licensee's inadequate corrective actions for repetitive failures of a TS required surveillance of the plant vent noble gas effluent monitor. Since July 2002, the monitor has failed the surveillance test five of the six times performed. The performance deficiency associated with this findings was inadequate corrective actions for repetitive failures of a TS required surveillance. This finding is more than minor significance because the R-27 radiation monitor was removed from service for periods in excess of twenty-four hours as a result of inadequate corrective actions. The finding was evaluated using the EP SDP, and was screened to be of very low safety significance, because there were alternate monitoring methods available in the event of an accident.
Inspection Report# : 2004003(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE EOF UPSs ON AUGUST 14 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the August 14, 2003, loss of off-site power event which revealed that Entergy did not have a preventive maintenance program in place to ensure the continued functionality of the numerous un-interruptible power supplies in the Emergency Operations Facility (EOF) which provide back-up power to emergency response equipment.
This finding is considered greater than minor because a significant amount of the Unit 2 and Unit 3 emergency response organization communications equipment was non-functional on August 14 until off-site power was restored. However, this finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions from the EOF, TSC/OSC, and Unit 2 and 3 central control rooms, using back-up telephone communications.
Inspection Report# : 2003013(pdf)
 
3Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 6 of 6 Significance:        Nov 07, 2003 Identified By: NRC Item Type: FIN Finding FAILURE OF THE UNIT 2 TSC DIESEL ON AUGUST 14, 2003 - FAILURE TO IMPLEMENT NON-RISK SIGNIFICANT PLANNING STANDARD PROGRAM ELEMENT This team-identified finding involves the failure of the Unit 2 TSC back-up diesel generator to function on August 14, 2003. The conditions which caused the diesel generator to fail to function involved electrical loading of the diesel generator in excess of its design capacity. This condition was initially identified in February 2000 and not resolved in a timely manner.
This finding is considered more than minor because a significant amount of TSC/OSC emergency response equipment, necessary to implement the Emergency Plan, was either de-energized by the Entergy staff because of the loss of sufficient air conditioning to ensure emergency response equipment would not be damaged due to overheating, or was without AC power because the diesel was non-functional. This finding is of very low safety significance because key members of the ERO were able to implement established compensatory measures to effectively perform their emergency response functions.
Inspection Report# : 2003013(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Dec 12, 2003 Identified By: NRC Item Type: FIN Finding The licensee's Corrective Action Program, used for identifying, tracking, prioritizing, and resolving deficiencies, was appropriately implemented in most cases.
Problem Identification & Resolution Team Inspection Summary: The inspection team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program, evaluating and prioritizing issues, and implementing appropriate corrective actions. However, the inspectors identified some minor equipment problems that had not been identified in the corrective action program. The evaluation of problems was generally adequate, but the inspectors identified two Green findings related to the failure to implement effective corrective acitons for degraded emergency preparedness equipment. These findings were determined not to involve violations of NRC requirements, however, they represented additional examples of the substantive cross-cutting issue in the problem identification and resolution area that was identified during the previous assessments. Based on interviews conducted during the inspection, station personnel felt free to identify safety issues and enter them into the corrective action program.
Inspection Report# : 2004003(pdf)
Last modified : December 29, 2004
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                              Page 1 of 7 Indian Point 2 4Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS ASSOCIATED WITH STATOR WATER COOLING PRESSURE SWITCH The inspector identified a self-revealing Green finding involving poor causal analysis associated with the main generator stator water cooling (SWC) system. The ineffective causal analysis was associated with the settings of the generator protection trip pressure switch (63-P79). The finding resulted in an automatic reactor trip due to a low inlet pressure condition on the main generator SWC system.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions, and is associated with the equipment performance attribute. Specifically, the finding affects the likelihood of a reactor trip and challenges the critical safety function of auxiliary feedwater (AFW) initiation. The finding is of very low risk significance (Green) since it does not contribute to both the likelihood of a reactor trip and the likelihood of mitigation equipment functions being unavailable. The finding is associated with the cross-cutting area of problem identification and resolution (PI&R) based on the ineffective causal analysis for previously identified deficiencies affecting the SWC system.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER INSTALLATION OF REACTOR COOLANT SYSTEM LOOP FLOW TUBING RESULTING IN REACTOR COOLANT SYSTEM LEAKAGE The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V "Instructions, Procedures and Drawings." Maintenance personnel did not verify that the length of tubing between the RACK 20 bulkhead connection and the existing 21 Reactor Coolant Loop Flow (FT-416) Hi side impulse tubing was sufficient for a proper Swagelok connection pursuant to procedure IP-SMM-MA-108.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenges critical safety functions, and is associated with the procedural quality attribute. Specifically, the finding affects the likelihood of a reactor coolant system (RCS) leak that upsets plant stability and challenges critical safety functions. This finding is of very low risk significance (Green) since worst case degradation would not result in exceeding the technical specification (TS) limit for identified leakage (10 gpm) and it does not affect the mitigation system's safety functions.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW RCS DRAINDOWN PROCEDURE DUE TO INAPPROPIATE APPROACH The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings" associated with a reactor vessel water level control issue during the drain down for the reactor vessel head re-installation on November 11, 2004. Specifically, an inappropriate level reduction rate existed by procedure, such that when communications to field operational personnel were temporarily lost and manual valve manipulations to reduce the rate were delayed, a two foot lower reactor vessel water level resulted.
This finding is more than minor, because it potentially affects the Initiating Events cornerstone objective of limiting the likelihood of events that challenge critical safety functions during shutdown, and is associated with the procedural quality attribute. This finding is considered to be of very low safety significance (Green), because residual heat removal (RHR) shutdown cooling remained operable and gravity re-flood of the reactor without operator action would have limited the consequences of any potential loss of shutdown cooling.
Inspection Report# : 2004012(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: FIN Finding
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 2 of 7 INADEQUATE CAUSAL ANALYSIS FOR 22 FEEDWATER REGULATING VALVE The inspectors identified a finding involving ineffective causal analysis for feedwater flow perturbations that led to a manual reactor trip on September 1, 2004. Ineffective causal analysis between September 1 - 5, resulted in two power escalation attempts without successfully identifying the direct cause of the feedwater flow perturbations. The effectiveness of Entergy's causal analysis was affected by informal troubleshooting and a variety of corrected equipment problems that did not support the underlying direct cause of the feedwater flow problem.
This finding is more than minor since if left uncorrected the finding would become a more significant safety concern. Specifically, if the effectiveness of Entergy's approach to causal analysis were not addressed, recurring plant transients and safety system challenges would result in a more significant safety concern. This finding affects the Initiating Event cornerstone since the two subsequent power changes did increase the likelihood of a reactor trip due to challenging reactor protection system (RPS) set points on steam generator level. The issue is considered to be of very low safety significance since the finding did not impact mitigation equipment availability or function. This issue was placed in Entergy's corrective action program (CAP) as CR-IP2-2004-04291. This finding is considered relevant to problem identification and resolution (PI&R) since it relates to Entergy's effectiveness in resolving problems.
Inspection Report# : 2004008(pdf)
Significance:        Sep 30, 2004 Identified By: Self Disclosing Item Type: FIN Finding FAILURE TO PROMPTLY IDENTIFY DEGRADED CONDITIONS ASSOCIATED WITH THE 23 FEEDWATER REGULATING VALVE A self-revealing Green finding related to the failure to promptly identify a degraded condition between September 2 - September 24 associated with the 23 feedwater regulating valve (FWRV) solenoid SOV-E. The failure to promptly identify and correct deficiencies associated with SOV-E resulted in a manual reactor trip on September 24, 2004. Entergy's actions were ineffective in that feedwater (FW) piping walkdowns following several feedwater transients failed to identify degradation of the solenoids' L-shaped conduit bracket. Furthermore, on September 20, 2004, when degradation of the L-shaped bracket for SOV-E was identified, it was not entered in Entergy's CAP. Subsequently, the degraded L-shaped bracket for SOV-E led to a manual reactor trip on September 24.
This finding was greater than minor since it adversely affected the Initiating Events cornerstone objective of limiting the likelihood of those events that upset plant stability (manual reactor trip) and challenge critical safety functions (initiation of auxiliary feedwater due to a partial loss of main FW flow) during power operations. The finding was associated with the cornerstone attribute of equipment performance since the solenoid valve for the 23 FWRV impacted the reliability of an FW isolation signal. The finding is of very low safety significance because the failure of the FW isolation solenoid contributed to the likelihood of a reactor trip; however, it did not affect the likelihood that other mitigation systems would not be available. On September 24, 2004, this issue was placed in Entergy's CAP as CR-IP2-2004-04522. This finding is considered relevant to PI&R since it relates to Entergy's effectiveness in identifying problems.
Inspection Report# : 2004008(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF TS 5.4.1.d FOR THE FAILURE TO PERFORM A TRANSIENT COMBUSTIBLE EVALUATION FOR 330 GALLONS OF OIL TEMPORARILY STORED IN FIRE ZONE 6A The inspector identified a non-cited violation of Technical Specification (TS) 5.4.1.d. that requires, in part, that written procedures shall be implemented for the Fire Protection Program. The inspector determined that no transient combustible evaluation (TCE) was completed for approximately 330 gallons of lubricating oil stored in fire zone 6A, "Waste Drumming and Storage Station," contrary to Procedure ENN-DC-161, "Transient Combustible Program," step 5.2.3.
This finding is greater than minor because it represented a condition similar to example 4.k in Appendix E, IMC 0612, in that the as-found condition involved transient combustible material loading in excess of the Fire Hazard Analysis limit. The finding is of very low safety significance because it did not increase the likelihood of a fire, no credible fire scenario was identified due to the type of storage containers used, there were no intervening combustibles, and no credible fire ignition source was present.
Inspection Report# : 2004002(pdf)
Mitigating Systems Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE PROCEDURES FOR EMERGENCY CORE COOLING SYSTEMS OPERATIONS The inspector identified a Green non-cited violation of TS 5.4.1 associated with Entergy's failure to properly implement procedure 2-COL 10.0,
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 3 of 7 "Locked Safeguards Valves." Residual heat removal recirculation valve AC-1863 was left in the shut position during the restart from IP2 refueling outage No. 16 (2RF16). The valve was not locked open in accordance with 2-COL 10.0 prior to entering Mode 4 due to the sequence of procedures performed at the end of the refueling outage.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of configuration control and adversely affects the capability of systems that respond to initiating events to prevent undesirable consequences. The finding involves the unavailability of a design feature described in the Final Safety Analysis Report (FSAR) that would ensure the capability to continue high-head recirculation after a loss of coolant accident (LOCA) in the event of certain system failures. This finding is of very low safety significance (Green), because the normal flow paths for establishing flow to the safety injection (SI) pump suctions during high-head recirculation remained available for the duration of the period that valve AC-1863 was shut. This finding is associated with the cross-cutting area of human performance, in that, operators did not adequately assess a change in the sequence of procedures performed during the refueling outage.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE PROCEDURE RESULTING IN ALL EDG'S BEING DECLARED INOPERABLE DUE TO DEFEATING SBO LOGIC The inspector identified a self-revealing Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings." A maintenance procedure for trip checks associated with the 345KV electrical feeder was inadequate since it did not provide appropriate directions for the test set-up. As a result, technicians unintentionally reset the main generator lock-out relays by using test stabs which defeated the station blackout (SBO) relays associated with the emergency diesel generators (EDGs) starting logic.
The finding is more than minor since it affects the procedure quality attribute of the Mitigating Systems cornerstone and impacts the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance (Green) due to low exposure time, credit for manual actions in the abnormal operating procedures (AOPs) to restore power to the safety-related 480 volt buses and start the required loads to stabilize plant conditions, and the availability of other mitigating equipment (ie. steam driven AFW pump and gas turbines 1 and 2).
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE PREVENTATIVE MAINTENANCE PROCEDURE IMPLEMENTATION RESULTING IN A LOSS OF SAFEGUARDS BUS 6A The inspector identified a self-revealing Green, non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding involved improper maintenance on a 480 volt cross-tie breaker (52/3AT6A). Maintenance personnel did not install the main line contactors for breaker (52/3AT6A) consistent with maintenance procedure BRK-P-003-A, "Westinghouse Model DB-75 Breaker -
Preventative Maintenance."
The finding is greater than minor since it affects the Mitigating Systems cornerstone objective of ensuring the availability of the RHR system and to prevent undesirable consequences such as core damage due to lack of core cooling during plant shutdown. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality (breaker preventative maintenance (PM) procedure). This finding is considered to be of very low safety significance since it did not degrade Entergy's ability to terminate a leak path or add reactor coolant inventory when needed, or degrade Entergy's ability to recover RHR once it is was lost. This finding is associated with the cross-cutting area of human performance, in that maintenance personnel did not implement a 480 volt breaker PM procedure correctly.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation MULTIPLE DEFICIENCIES IN SURVEILLANCE PROCEDURES ASSOCIATED WITH ITS CONVERSION The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criterion VI, "Document Control." Inadequate document control resulted in multiple surveillance procedures not meeting the criteria of the Improved Technical Specifications (ITS) surveillance requirements (SRs) or the applicable ITS basis document.
The finding is more than minor since, if left uncorrected, it would become a more significant safety concern potentially impacting multiple SRs of safety-related equipment and equipment important to safety. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality. This finding is considered to be of very low risk significance (Green) since it had not resulted in a loss of safety function or in any inoperable equipment.
Inspection Report# : 2004012(pdf)
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                        Page 4 of 7 Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate and timely corrective actions for known deficiencies in the control program(s) and installation of safety related electrical cables and raceways.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address conditions adverse to quality concerning one example of resolution of Data Verification Transfer Report (DVTR) Items/Operability Assessments; and two examples of configuration control of electrical raceways and cables.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls for electrical cable and raceway installations.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for Entergy's failure to implement appropriate design controls for the installation of safety related electrical cables and raceways.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly control cable separation program documents.
Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for Entergy's failure to properly control the cable separation program documents. These documents include some reports never being reviewed, approved, and signed off as well as documents used in part for design specifications and DBD work not entered into the document control program to ensure retrieveability.
Inspection Report# : 2004009(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a recirculation sump screen bypass path.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality. Specifically, Entergy did not promptly identify and correct a recirculation sump bypass path and containment debris that had the potential to adversely impact ECCS during containment recirculation.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a Technical Specification Surveillance Requirement The inspector identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.3.1.1. that requires, in part, that a channel check be performed every 12 hours on the feedwater flow instrumentation in the central control room. This requirement had not been met since the licensee implemented the Improved Technical Specifications in December of 2003.
This finding is greater than minor because it represented a condition similar to example 1.c in Appendix E, IMC 0612, in that the surveillance was not performed per Technical Specifications from December 12, 2003 through June 8, 2004. The finding is of very low safety significance because the feedwater flow instruments met the surveillance criteria when subsequently performed, and did not render the mitigating equipment inoperable.
Inspection Report# : 2004006(pdf)
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                          Page 5 of 7 Significance:        Jun 30, 2004 Identified By: NRC Item Type: FIN Finding Failure to implement adequate corrective actions for low voltage conditions on the 13.8 kV system.
The inspectors identified a finding due to ineffective and untimely corrective actions associated with the 13.8 KV system during reduced voltage conditions.
This finding was determined to be greater than minor since it impacts the mitigating systems cornerstone objective of ensuring system reliability and capability. This finding was associated with the procedure quality attribute of that cornerstone. This finding was of very low safety significance since there was no loss of the normal offsite power supplies and the 13.8 KV system was not providing power to any safety-related loads during the degraded condition.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls during modifications to the recirculation sump.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for Entergy's failure to translate the emergency core cooling system (ECCS) design basis into recirculation sump modification instructions. Specifically, Entergy added penetration cover plates and alignment collars around several small pipes that penetrated the sump deck plating, and the annular gap between the collars and pipes exceeded the sump screen size.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (loss-of-coolant accidents) (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a condition adverse to quality which could impact EDG reliability.
The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality concerning emergency diesel generator (EDG) heat exchanger (HX) fouling.
This finding was more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring equipment availability and reliability of the EDG HXs to perform their intended safety function. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone. However, this finding was determined to have very low safety significance (Green) using the SDP Phase 1 screening worksheet for mitigating systems because the EDG HXs remained operable and capable of performing their intended safety function.
Inspection Report# : 2004006(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE FAILURE TO TAKE APPROPRIATE CORRECTIVE ACTIONS TO ENSURE THE RELIABILITY AND AVAILABILITY OF GAS TURBINE NO. 1 A finding was identified involving untimely corrective actions which contributed to increased unavailability of Gas Turbine 1 (GT-1) which is considered a mitigating system. Specifically, GT-1 was not available for approximately 116 hours due to the failure and subsequent replacement of the starting diesel battery charger.
This finding was determined to be greater than minor since it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring system reliability and availability of systems that are used to prevent undesirable consequences due to initiating events. GT-1 is credited as an alternate AC power source for both station blackout and Appendix R fire scenarios. This finding was considered of very low safety significance because there was no actual loss of safety function for this mitigating system, since GT-3 was available while GT-1 was inoperable. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                            Page 6 of 7 Item Type: FIN Finding A VERY LOW RISK SIGNIFICANT FINDING INVOLVING THE IMPROPER CONTROL OF AN OUT-GOING 345 KV FEEDER BREAKER DURING A FEEDER OUTAGE The inspector identified that the control room operators placed the 345KV ring bus in a configuration that would challenge the availability of mitigating systems in the event of an off-site electrical transient. Specifically, in the event that a 345KV feeder fault caused a loss-of-load plant trip, two of the four 480 volt safety buses would require operator action to restore.
This finding is greater than minor since it is associated with the configuration control attribute of the mitigating systems cornerstone and that it impacted the cornerstone objective of ensuring the availability of mitigating systems. With the ring bus aligned with both output breakers shut and one feeder out of service, a subsequent fault on the remaining feeder would have resulted in a plant trip with the fast transfer blocked. This would de-energize vital busses 2A and 3A causing a loss of power to one of two motor-driven auxiliary feed pumps, one of three safety injection pumps, two of five containment fan cooler units, one of two residual heat removal pumps and two of six service water pumps. Manual operator action would then be required to restore this equipment. This finding is considered of very low safety significance because it did not result in an actual loss of safety function of any mitigating systems. This issue did not screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF 10 CFR 50, APPENDIX B, CRITERION III "DESIGN CONTROL" FOR THE FAILURE TO IMPLEMENT APPROPRIATE DESIGN CONTROLS FOR A MODIFICATION MADE TO THE OTDT CONTROLLER The inspector identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control." A component was modified during the replacement of a safety-related controller in the over-temperature delta-temperature (OTDT) circuitry of the reactor protection system without a formal modification package.
This finding was determined to be greater than minor since it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring reactor protection system reliability. Specifically, the failure to use a derated resistor to modify the circuit card had an adverse impact on the reliability of the OTDT controller. This finding is considered of very low safety significance since it did not result in the actual loss of safety function of a system. This issue did not impact fire, flooding, seismic, or severe weather initiating events.
Inspection Report# : 2004002(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO CONDUCT SIMULATOR TESTING IN ACCORDANCE WITH ANSI/ANS 3.5-1985 The inspectors identified that simulator performance testing did not meet the standards as specified in ANSI/ANS 3.5-1985 in that: (1) "best estimate" data for the simulator testing was not used; (2) all required key parameters during the simulator test were not recorded; and (3) simulator differences identified during testing were not documented and justified.
This finding is more than minor because it affects the human performance (human error) attribute of the mitigating systems cornerstone. More specifically, improperly conducted simulator testing resulted, in part, in not identifying replication issues for steam generator pressure and cold leg temperature. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
Significance:        Jan 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE SIMULATOR TO DEMONSTRATE EXPECTED PLANT RESPONSE TO TRANSIENT CONDITIONS The inspectors identified a non-cited violation of 10 CFR 55.46(c)(1), involving the failure of the simulator to correctly replicate key parameters such as steam generator pressure and cold leg temperature (Tcold) during a loss of all reactor coolant pumps. Additionally, the plant decay heat load was not correctly modeled which contributed to inappropriate operator actions during the August 3, 2003, plant trip.
This finding is more than minor because it affected the human performance (human error) attribute of the mitigating systems cornerstone. Not correctly replicating the plant's response on the simulator provides the potential for negative operator training. The finding is of very low safety significance (Green) because the discrepancy did not have an adverse impact on operator actions such that safety related equipment was made inoperable during normal operations or in response to a plant transient.
Inspection Report# : 2004004(pdf)
 
4Q/2004 Inspection Findings - Indian Point 2                                                                                        Page 7 of 7 Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY ADDRESS A CONDITION ADVERSE TO QUALITY INVOLVING LEAKAGE FROM A CANOPY SEAL WELD ONTO THE REACTOR VESSEL HEAD IN NOVEMBER 2002 The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for Entergy's failure to properly address a condition adverse to quality involving leakage from a canopy seal weld in November 2002. The ineffective corrective actions for this conoseal leak led to boron accumulation on the reactor vessel head (RVH).
The finding is considered to be more than minor since, if left uncorrected, it could have led to a more significant problem. Specifically, the boric acid, if re-wetted, could have led to accelerated corrosion of the RVH. The finding is of very low significance since the RVH integrity was not affected by this problem. The finding is associated with the cross-cutting area of PI&R related to the ineffective corrective actions for the conoseal leak.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE INSPECTION CRITERIA AND GUIDANCE TO EVALUATORS PRIOR TO THE INSPECTION OF THE REACTOR VESSEL LOWER HEAD PENETRATION NOZZLES The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion IX, "Control of Special Processes," for Entergy's failure to provide adequate inspection criteria and guidance to evaluators prior to the inspection of the reactor vessel lower head penetration nozzles. In particular, Entergy personnel performed visual inspections of the reactor vessel bottom mounted instrumentation annulus area without adequate procedural guidance to define potential problems or indications.
This finding is considered to be more than minor since inspection program deficiencies could allow a degraded component to remain inservice undetected. Specifically, the failure to develop adequate inspection guidance could result in a failure to detect a degraded lower RVH penetration boundary. The finding is of very low significance since the lower RVH integrity was not affected.
Inspection Report# : 2004012(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 09, 2005
 
1Q/2005 Inspection Findings - Indian Point 2                                                                                              Page 1 of 6 Indian Point 2 1Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS ASSOCIATED WITH STATOR WATER COOLING PRESSURE SWITCH The inspector identified a self-revealing Green finding involving poor causal analysis associated with the main generator stator water cooling (SWC) system. The ineffective causal analysis was associated with the settings of the generator protection trip pressure switch (63-P79). The finding resulted in an automatic reactor trip due to a low inlet pressure condition on the main generator SWC system.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions, and is associated with the equipment performance attribute. Specifically, the finding affects the likelihood of a reactor trip and challenges the critical safety function of auxiliary feedwater (AFW) initiation. The finding is of very low risk significance (Green) since it does not contribute to both the likelihood of a reactor trip and the likelihood of mitigation equipment functions being unavailable. The finding is associated with the cross-cutting area of problem identification and resolution (PI&R) based on the ineffective causal analysis for previously identified deficiencies affecting the SWC system.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER INSTALLATION OF REACTOR COOLANT SYSTEM LOOP FLOW TUBING RESULTING IN REACTOR COOLANT SYSTEM LEAKAGE The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V "Instructions, Procedures and Drawings." Maintenance personnel did not verify that the length of tubing between the RACK 20 bulkhead connection and the existing 21 Reactor Coolant Loop Flow (FT-416) Hi side impulse tubing was sufficient for a proper Swagelok connection pursuant to procedure IP-SMM-MA-108.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenges critical safety functions, and is associated with the procedural quality attribute. Specifically, the finding affects the likelihood of a reactor coolant system (RCS) leak that upsets plant stability and challenges critical safety functions. This finding is of very low risk significance (Green) since worst case degradation would not result in exceeding the technical specification (TS) limit for identified leakage (10 gpm) and it does not affect the mitigation system's safety functions.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW RCS DRAINDOWN PROCEDURE DUE TO INAPPROPIATE APPROACH The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings" associated with a reactor vessel water level control issue during the drain down for the reactor vessel head re-installation on November 11, 2004. Specifically, an inappropriate level reduction rate existed by procedure, such that when communications to field operational personnel were temporarily lost and manual valve manipulations to reduce the rate were delayed, a two foot lower reactor vessel water level resulted.
This finding is more than minor, because it potentially affects the Initiating Events cornerstone objective of limiting the likelihood of events that challenge critical safety functions during shutdown, and is associated with the procedural quality attribute. This finding is considered to be of very low safety significance (Green), because residual heat removal (RHR) shutdown cooling remained operable and gravity re-flood of the reactor without operator action would have limited the consequences of any potential loss of shutdown cooling.
Inspection Report# : 2004012(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: FIN Finding
 
1Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 2 of 6 INADEQUATE CAUSAL ANALYSIS FOR 22 FEEDWATER REGULATING VALVE The inspectors identified a finding involving ineffective causal analysis for feedwater flow perturbations that led to a manual reactor trip on September 1, 2004. Ineffective causal analysis between September 1 - 5, resulted in two power escalation attempts without successfully identifying the direct cause of the feedwater flow perturbations. The effectiveness of Entergy's causal analysis was affected by informal troubleshooting and a variety of corrected equipment problems that did not support the underlying direct cause of the feedwater flow problem.
This finding is more than minor since if left uncorrected the finding would become a more significant safety concern. Specifically, if the effectiveness of Entergy's approach to causal analysis were not addressed, recurring plant transients and safety system challenges would result in a more significant safety concern. This finding affects the Initiating Event cornerstone since the two subsequent power changes did increase the likelihood of a reactor trip due to challenging reactor protection system (RPS) set points on steam generator level. The issue is considered to be of very low safety significance since the finding did not impact mitigation equipment availability or function. This issue was placed in Entergy's corrective action program (CAP) as CR-IP2-2004-04291. This finding is considered relevant to problem identification and resolution (PI&R) since it relates to Entergy's effectiveness in resolving problems.
Inspection Report# : 2004008(pdf)
Significance:        Sep 30, 2004 Identified By: Self Disclosing Item Type: FIN Finding FAILURE TO PROMPTLY IDENTIFY DEGRADED CONDITIONS ASSOCIATED WITH THE 23 FEEDWATER REGULATING VALVE A self-revealing Green finding related to the failure to promptly identify a degraded condition between September 2 - September 24 associated with the 23 feedwater regulating valve (FWRV) solenoid SOV-E. The failure to promptly identify and correct deficiencies associated with SOV-E resulted in a manual reactor trip on September 24, 2004. Entergy's actions were ineffective in that feedwater (FW) piping walkdowns following several feedwater transients failed to identify degradation of the solenoids' L-shaped conduit bracket. Furthermore, on September 20, 2004, when degradation of the L-shaped bracket for SOV-E was identified, it was not entered in Entergy's CAP. Subsequently, the degraded L-shaped bracket for SOV-E led to a manual reactor trip on September 24.
This finding was greater than minor since it adversely affected the Initiating Events cornerstone objective of limiting the likelihood of those events that upset plant stability (manual reactor trip) and challenge critical safety functions (initiation of auxiliary feedwater due to a partial loss of main FW flow) during power operations. The finding was associated with the cornerstone attribute of equipment performance since the solenoid valve for the 23 FWRV impacted the reliability of an FW isolation signal. The finding is of very low safety significance because the failure of the FW isolation solenoid contributed to the likelihood of a reactor trip; however, it did not affect the likelihood that other mitigation systems would not be available. On September 24, 2004, this issue was placed in Entergy's CAP as CR-IP2-2004-04522. This finding is considered relevant to PI&R since it relates to Entergy's effectiveness in identifying problems.
Inspection Report# : 2004008(pdf)
Mitigating Systems Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE PROCEDURES FOR EMERGENCY CORE COOLING SYSTEMS OPERATIONS The inspector identified a Green non-cited violation of TS 5.4.1 associated with Entergy's failure to properly implement procedure 2-COL 10.0, "Locked Safeguards Valves." Residual heat removal recirculation valve AC-1863 was left in the shut position during the restart from IP2 refueling outage No. 16 (2RF16). The valve was not locked open in accordance with 2-COL 10.0 prior to entering Mode 4 due to the sequence of procedures performed at the end of the refueling outage.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of configuration control and adversely affects the capability of systems that respond to initiating events to prevent undesirable consequences. The finding involves the unavailability of a design feature described in the Final Safety Analysis Report (FSAR) that would ensure the capability to continue high-head recirculation after a loss of coolant accident (LOCA) in the event of certain system failures. This finding is of very low safety significance (Green), because the normal flow paths for establishing flow to the safety injection (SI) pump suctions during high-head recirculation remained available for the duration of the period that valve AC-1863 was shut. This finding is associated with the cross-cutting area of human performance, in that, operators did not adequately assess a change in the sequence of procedures performed during the refueling outage.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation
 
1Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 3 of 6 INADEQUATE MAINTENANCE PROCEDURE RESULTING IN ALL EDG'S BEING DECLARED INOPERABLE DUE TO DEFEATING SBO LOGIC The inspector identified a self-revealing Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings." A maintenance procedure for trip checks associated with the 345KV electrical feeder was inadequate since it did not provide appropriate directions for the test set-up. As a result, technicians unintentionally reset the main generator lock-out relays by using test stabs which defeated the station blackout (SBO) relays associated with the emergency diesel generators (EDGs) starting logic.
The finding is more than minor since it affects the procedure quality attribute of the Mitigating Systems cornerstone and impacts the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance (Green) due to low exposure time, credit for manual actions in the abnormal operating procedures (AOPs) to restore power to the safety-related 480 volt buses and start the required loads to stabilize plant conditions, and the availability of other mitigating equipment (ie. steam driven AFW pump and gas turbines 1 and 2).
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE PREVENTATIVE MAINTENANCE PROCEDURE IMPLEMENTATION RESULTING IN A LOSS OF SAFEGUARDS BUS 6A The inspector identified a self-revealing Green, non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding involved improper maintenance on a 480 volt cross-tie breaker (52/3AT6A). Maintenance personnel did not install the main line contactors for breaker (52/3AT6A) consistent with maintenance procedure BRK-P-003-A, "Westinghouse Model DB-75 Breaker -
Preventative Maintenance."
The finding is greater than minor since it affects the Mitigating Systems cornerstone objective of ensuring the availability of the RHR system and to prevent undesirable consequences such as core damage due to lack of core cooling during plant shutdown. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality (breaker preventative maintenance (PM) procedure). This finding is considered to be of very low safety significance since it did not degrade Entergy's ability to terminate a leak path or add reactor coolant inventory when needed, or degrade Entergy's ability to recover RHR once it is was lost. This finding is associated with the cross-cutting area of human performance, in that maintenance personnel did not implement a 480 volt breaker PM procedure correctly.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation MULTIPLE DEFICIENCIES IN SURVEILLANCE PROCEDURES ASSOCIATED WITH ITS CONVERSION The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criterion VI, "Document Control." Inadequate document control resulted in multiple surveillance procedures not meeting the criteria of the Improved Technical Specifications (ITS) surveillance requirements (SRs) or the applicable ITS basis document.
The finding is more than minor since, if left uncorrected, it would become a more significant safety concern potentially impacting multiple SRs of safety-related equipment and equipment important to safety. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality. This finding is considered to be of very low risk significance (Green) since it had not resulted in a loss of safety function or in any inoperable equipment.
Inspection Report# : 2004012(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate and timely corrective actions for known deficiencies in the control program(s) and installation of safety related electrical cables and raceways.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address conditions adverse to quality concerning one example of resolution of Data Verification Transfer Report (DVTR) Items/Operability Assessments; and two examples of configuration control of electrical raceways and cables.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 4 of 6 Failure to implement appropriate design controls for electrical cable and raceway installations.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for Entergy's failure to implement appropriate design controls for the installation of safety related electrical cables and raceways.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly control cable separation program documents.
Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for Entergy's failure to properly control the cable separation program documents. These documents include some reports never being reviewed, approved, and signed off as well as documents used in part for design specifications and DBD work not entered into the document control program to ensure retrieveability.
Inspection Report# : 2004009(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: FIN Finding Failure to implement adequate corrective actions for low voltage conditions on the 13.8 kV system.
The inspectors identified a finding due to ineffective and untimely corrective actions associated with the 13.8 KV system during reduced voltage conditions.
This finding was determined to be greater than minor since it impacts the mitigating systems cornerstone objective of ensuring system reliability and capability. This finding was associated with the procedure quality attribute of that cornerstone. This finding was of very low safety significance since there was no loss of the normal offsite power supplies and the 13.8 KV system was not providing power to any safety-related loads during the degraded condition.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls during modifications to the recirculation sump.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for Entergy's failure to translate the emergency core cooling system (ECCS) design basis into recirculation sump modification instructions. Specifically, Entergy added penetration cover plates and alignment collars around several small pipes that penetrated the sump deck plating, and the annular gap between the collars and pipes exceeded the sump screen size.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (loss-of-coolant accidents) (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a recirculation sump screen bypass path.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality. Specifically, Entergy did not promptly identify and correct a recirculation sump bypass path and containment debris that had the potential to adversely impact ECCS during containment recirculation.
This finding is more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of ECCS to respond to initiating events (LOCAs) to prevent undesirable conditions. This finding is considered to be of very low safety significance, because ECCS remained operable and there was no loss of safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2005 Inspection Findings - Indian Point 2                                                                                        Page 5 of 6 Failure to identify a condition adverse to quality which could impact EDG reliability.
The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly identify and take actions to address a condition adverse to quality concerning emergency diesel generator (EDG) heat exchanger (HX) fouling.
This finding was more than minor because it potentially affected the mitigating systems cornerstone objective of ensuring equipment availability and reliability of the EDG HXs to perform their intended safety function. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone. However, this finding was determined to have very low safety significance (Green) using the SDP Phase 1 screening worksheet for mitigating systems because the EDG HXs remained operable and capable of performing their intended safety function.
Inspection Report# : 2004006(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a Technical Specification Surveillance Requirement The inspector identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.3.1.1. that requires, in part, that a channel check be performed every 12 hours on the feedwater flow instrumentation in the central control room. This requirement had not been met since the licensee implemented the Improved Technical Specifications in December of 2003.
This finding is greater than minor because it represented a condition similar to example 1.c in Appendix E, IMC 0612, in that the surveillance was not performed per Technical Specifications from December 12, 2003 through June 8, 2004. The finding is of very low safety significance because the feedwater flow instruments met the surveillance criteria when subsequently performed, and did not render the mitigating equipment inoperable.
Inspection Report# : 2004006(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY ADDRESS A CONDITION ADVERSE TO QUALITY INVOLVING LEAKAGE FROM A CANOPY SEAL WELD ONTO THE REACTOR VESSEL HEAD IN NOVEMBER 2002 The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for Entergy's failure to properly address a condition adverse to quality involving leakage from a canopy seal weld in November 2002. The ineffective corrective actions for this conoseal leak led to boron accumulation on the reactor vessel head (RVH).
The finding is considered to be more than minor since, if left uncorrected, it could have led to a more significant problem. Specifically, the boric acid, if re-wetted, could have led to accelerated corrosion of the RVH. The finding is of very low significance since the RVH integrity was not affected by this problem. The finding is associated with the cross-cutting area of PI&R related to the ineffective corrective actions for the conoseal leak.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE INSPECTION CRITERIA AND GUIDANCE TO EVALUATORS PRIOR TO THE INSPECTION OF THE REACTOR VESSEL LOWER HEAD PENETRATION NOZZLES The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion IX, "Control of Special Processes," for Entergy's failure to provide adequate inspection criteria and guidance to evaluators prior to the inspection of the reactor vessel lower head penetration nozzles. In particular, Entergy personnel performed visual inspections of the reactor vessel bottom mounted instrumentation annulus area without adequate procedural guidance to define potential problems or indications.
This finding is considered to be more than minor since inspection program deficiencies could allow a degraded component to remain inservice undetected. Specifically, the failure to develop adequate inspection guidance could result in a failure to detect a degraded lower RVH penetration boundary. The finding is of very low significance since the lower RVH integrity was not affected.
Inspection Report# : 2004012(pdf)
 
1Q/2005 Inspection Findings - Indian Point 2            Page 6 of 6 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : June 17, 2005
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                              Page 1 of 7 Indian Point 2 2Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding INADEQUATE CAUSAL ANALYSIS ASSOCIATED WITH STATOR WATER COOLING PRESSURE SWITCH The inspector identified a self-revealing Green finding involving poor causal analysis associated with the main generator stator water cooling (SWC) system. The ineffective causal analysis was associated with the settings of the generator protection trip pressure switch (63-P79). The finding resulted in an automatic reactor trip due to a low inlet pressure condition on the main generator SWC system.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions, and is associated with the equipment performance attribute. Specifically, the finding affects the likelihood of a reactor trip and challenges the critical safety function of auxiliary feedwater (AFW) initiation. The finding is of very low risk significance (Green) since it does not contribute to both the likelihood of a reactor trip and the likelihood of mitigation equipment functions being unavailable. The finding is associated with the cross-cutting area of problem identification and resolution (PI&R) based on the ineffective causal analysis for previously identified deficiencies affecting the SWC system.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation IMPROPER INSTALLATION OF REACTOR COOLANT SYSTEM LOOP FLOW TUBING RESULTING IN REACTOR COOLANT SYSTEM LEAKAGE The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V "Instructions, Procedures and Drawings." Maintenance personnel did not verify that the length of tubing between the RACK 20 bulkhead connection and the existing 21 Reactor Coolant Loop Flow (FT-416) Hi side impulse tubing was sufficient for a proper Swagelok connection pursuant to procedure IP-SMM-MA-108.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenges critical safety functions, and is associated with the procedural quality attribute. Specifically, the finding affects the likelihood of a reactor coolant system (RCS) leak that upsets plant stability and challenges critical safety functions. This finding is of very low risk significance (Green) since worst case degradation would not result in exceeding the technical specification (TS) limit for identified leakage (10 gpm) and it does not affect the mitigation system's safety functions.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW RCS DRAINDOWN PROCEDURE DUE TO INAPPROPIATE APPROACH The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings" associated with a reactor vessel water level control issue during the drain down for the reactor vessel head re-installation on November 11, 2004. Specifically, an inappropriate level reduction rate existed by procedure, such that when communications to field operational personnel were temporarily lost and manual valve manipulations to reduce the rate were delayed, a two foot lower reactor vessel water level resulted.
This finding is more than minor, because it potentially affects the Initiating Events cornerstone objective of limiting the likelihood of events that challenge critical safety functions during shutdown, and is associated with the procedural quality attribute. This finding is considered to be of very low safety significance (Green), because residual heat removal (RHR) shutdown cooling remained operable and gravity re-flood of the reactor without operator action would have limited the consequences of any potential loss of shutdown cooling.
Inspection Report# : 2004012(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: FIN Finding
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 2 of 7 INADEQUATE CAUSAL ANALYSIS FOR 22 FEEDWATER REGULATING VALVE The inspectors identified a finding involving ineffective causal analysis for feedwater flow perturbations that led to a manual reactor trip on September 1, 2004. Ineffective causal analysis between September 1 - 5, resulted in two power escalation attempts without successfully identifying the direct cause of the feedwater flow perturbations. The effectiveness of Entergy's causal analysis was affected by informal troubleshooting and a variety of corrected equipment problems that did not support the underlying direct cause of the feedwater flow problem.
This finding is more than minor since if left uncorrected the finding would become a more significant safety concern. Specifically, if the effectiveness of Entergy's approach to causal analysis were not addressed, recurring plant transients and safety system challenges would result in a more significant safety concern. This finding affects the Initiating Event cornerstone since the two subsequent power changes did increase the likelihood of a reactor trip due to challenging reactor protection system (RPS) set points on steam generator level. The issue is considered to be of very low safety significance since the finding did not impact mitigation equipment availability or function. This issue was placed in Entergy's corrective action program (CAP) as CR-IP2-2004-04291. This finding is considered relevant to problem identification and resolution (PI&R) since it relates to Entergy's effectiveness in resolving problems.
Inspection Report# : 2004008(pdf)
Significance:        Sep 30, 2004 Identified By: Self Disclosing Item Type: FIN Finding FAILURE TO PROMPTLY IDENTIFY DEGRADED CONDITIONS ASSOCIATED WITH THE 23 FEEDWATER REGULATING VALVE A self-revealing Green finding related to the failure to promptly identify a degraded condition between September 2 - September 24 associated with the 23 feedwater regulating valve (FWRV) solenoid SOV-E. The failure to promptly identify and correct deficiencies associated with SOV-E resulted in a manual reactor trip on September 24, 2004. Entergy's actions were ineffective in that feedwater (FW) piping walkdowns following several feedwater transients failed to identify degradation of the solenoids' L-shaped conduit bracket. Furthermore, on September 20, 2004, when degradation of the L-shaped bracket for SOV-E was identified, it was not entered in Entergy's CAP. Subsequently, the degraded L-shaped bracket for SOV-E led to a manual reactor trip on September 24.
This finding was greater than minor since it adversely affected the Initiating Events cornerstone objective of limiting the likelihood of those events that upset plant stability (manual reactor trip) and challenge critical safety functions (initiation of auxiliary feedwater due to a partial loss of main FW flow) during power operations. The finding was associated with the cornerstone attribute of equipment performance since the solenoid valve for the 23 FWRV impacted the reliability of an FW isolation signal. The finding is of very low safety significance because the failure of the FW isolation solenoid contributed to the likelihood of a reactor trip; however, it did not affect the likelihood that other mitigation systems would not be available. On September 24, 2004, this issue was placed in Entergy's CAP as CR-IP2-2004-04522. This finding is considered relevant to PI&R since it relates to Entergy's effectiveness in identifying problems.
Inspection Report# : 2004008(pdf)
Mitigating Systems Significance:        May 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE POST-ACCIDENT RECIRCULATION PUMP MOTOR LOADING CONDITIONS USED TO DETERMINE OVERLOAD TRIP SETTINGS FOR 480 VOLT TYPE DB CIRCUIT BREAKERS The team identified a finding where Entergy had used non-conservative post-accident recirculation pump motor loading conditions in an analysis that determined overload trip settings for the associated 480 Volt circuit breakers. This finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion III (Design Control).
This finding is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. This finding is of very low safety significance because it is a design deficiency that did not result in a loss of function.
Inspection Report# : 2005006(pdf)
Significance:        May 17, 2005 Identified By: NRC Item Type: AV Apparent Violation FAILURE TO ADEQUATELY EVALUATE AND CORRECT NITROGEN GAS MIGRATION AND ACCUMULATION IN PORTIONS OF THE SAFETY INJECTION SYSTEM An apparent violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Action) and station procedures were identified associated with the failure to evaluate and correct a condition adverse to quality. Specifically, the condition adverse to quality involved the leakage of water from
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 3 of 7 the No. 24 safety injection accumulator past several closed valves, allowing water containing absorbed nitrogen to reach other portions of the safety injection emergency core cooling system (including the common suction supply piping for the safety injection pumps and the 23 safety injection pump casing). As the water moved from a higher to lower system pressure, the nitrogen gas was released from the water, thereby challenging the performance of the safety injection pumps. In addition, Entergys initial evaluation of this condition did not appropriately consider available industry operating experience relative to gas migration into emergency core cooling system piping.
This issue is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The Significance Determination Process (SDP) Phase 1, Phase 2, and Phase 3 were used to determine that this issue represented a finding with preliminarily low to moderate safety significance. The analysis used the NRCs best functionality estimates for the three safety injection pumps over a 17-day period when it was judged that adverse gas accumulation conditions existed. Specifically, the 23 safety injection pump was not functional due to the pump casing being filled with gas. The team concluded that the 21 and 22 pumps, given the accumulated gas in the pump suction piping, would not have functioned 75% of the time (assigned a 75% failure probability) for high flowrate and low discharge pressure conditions in response to a medium break loss of coolant accident; and 25% of the time for low flowrate and high discharge pressure conditions in response to other initiating events. The Phase 1 screening identified that a Phase 2 analysis was needed because the 23 safety injection pump train was not functional for longer than the technical specification allowed outage time of 72 hours. Given the uncertainty in the Phase 2 analysis, a Phase 3 analysis was necessary to improve the accuracy of the result. The Phase 3 analysis for internal and external initiating events, using the above assumptions and licensee risk information, identified an increase in core damage frequency of approximately 1 in 900,000 years of operation (low E-6 per year range); and an increase in large early release frequency of approximately 1 in 3,000,000 years of operation (low E-7 per year range).
Inspection Report# : 2005006(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding FAILURE TO PERIODICALLY VERIFY THE CAPABILITY OF CITY WATER BACKUP COOLING SAFETY FUNCTION The inspectors identified a Green finding associated with a loss of city water to the primary auxiliary building on January 26, 2005.
Specifically, Entergy failed to periodically verify the capability of a backup cooling water supply for the charging pumps, safety injection pumps and the residual heat removal pumps.
The finding is greater than minor since it affected the Mitigating Systems cornerstone objective of availability of backup cooling to safety pumps in response to a loss of all component cooling water and/or loss of service water event. This finding impacted the procedural quality attribute since no periodic verification existed since 2003 to verify the availability of backup cooling water source, city water. In accordance with IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the Region I Senior Reactor Analyst (SRA) performed a Phase 3 analysis and determined that this finding was of very low risk significance (Green). No violations of NRC requirements were identified.
Inspection Report# : 2005002(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE INTERIM COMPENSATORY MEASURES FOR FIRE BARRIER IMPAIRMENTS The inspectors identified a Green non-cited violation of license condition 2.K between November 26, 2004 - March 9, 2005, due to inadequate compensatory actions for a degraded 3-hour rated fire barrier (3M Interam) for penetration H20 concurrent with a degraded hose station nearest to the fire barrier H20. Penetration H20 houses electrical cables needed for the Alternate Safe Shutdown System.
The finding is more than minor since, if left uncorrected, the finding would become a more significant safety concern. The finding affects the Mitigating Systems cornerstone, and its objective of ensuring availability, reliability and capability of systems that respond to initiating events, since both deficiencies contributed to plant risk by decreasing the endurance of the fire barrier and affecting the ability to manually (no automatic suppression capability) fight fires in the electrical penetration room. This issue was of very low risk significance (Green) using phase 1 of the Fire Protection SDP, MC 0612 Appendix F because the barrier was judged to afford greater than 20 minutes of fire endurance protection and low combustible loading was found in the fire area. This finding is associated with the cross-cutting area of human performance (personnel) in that fire protection engineering did not document or implement adequate compensatory measures for the degraded fire barrier and inoperable hose station.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE PROCEDURES FOR EMERGENCY CORE COOLING SYSTEMS OPERATIONS The inspector identified a Green non-cited violation of TS 5.4.1 associated with Entergy's failure to properly implement procedure 2-COL 10.0, "Locked Safeguards Valves." Residual heat removal recirculation valve AC-1863 was left in the shut position during the restart from IP2
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 4 of 7 refueling outage No. 16 (2RF16). The valve was not locked open in accordance with 2-COL 10.0 prior to entering Mode 4 due to the sequence of procedures performed at the end of the refueling outage.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of configuration control and adversely affects the capability of systems that respond to initiating events to prevent undesirable consequences. The finding involves the unavailability of a design feature described in the Final Safety Analysis Report (FSAR) that would ensure the capability to continue high-head recirculation after a loss of coolant accident (LOCA) in the event of certain system failures. This finding is of very low safety significance (Green), because the normal flow paths for establishing flow to the safety injection (SI) pump suctions during high-head recirculation remained available for the duration of the period that valve AC-1863 was shut. This finding is associated with the cross-cutting area of human performance, in that, operators did not adequately assess a change in the sequence of procedures performed during the refueling outage.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE PROCEDURE RESULTING IN ALL EDG'S BEING DECLARED INOPERABLE DUE TO DEFEATING SBO LOGIC The inspector identified a self-revealing Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings." A maintenance procedure for trip checks associated with the 345KV electrical feeder was inadequate since it did not provide appropriate directions for the test set-up. As a result, technicians unintentionally reset the main generator lock-out relays by using test stabs which defeated the station blackout (SBO) relays associated with the emergency diesel generators (EDGs) starting logic.
The finding is more than minor since it affects the procedure quality attribute of the Mitigating Systems cornerstone and impacts the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance (Green) due to low exposure time, credit for manual actions in the abnormal operating procedures (AOPs) to restore power to the safety-related 480 volt buses and start the required loads to stabilize plant conditions, and the availability of other mitigating equipment (ie. steam driven AFW pump and gas turbines 1 and 2).
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation INADEQUATE PREVENTATIVE MAINTENANCE PROCEDURE IMPLEMENTATION RESULTING IN A LOSS OF SAFEGUARDS BUS 6A The inspector identified a self-revealing Green, non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding involved improper maintenance on a 480 volt cross-tie breaker (52/3AT6A). Maintenance personnel did not install the main line contactors for breaker (52/3AT6A) consistent with maintenance procedure BRK-P-003-A, "Westinghouse Model DB-75 Breaker -
Preventative Maintenance."
The finding is greater than minor since it affects the Mitigating Systems cornerstone objective of ensuring the availability of the RHR system and to prevent undesirable consequences such as core damage due to lack of core cooling during plant shutdown. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality (breaker preventative maintenance (PM) procedure). This finding is considered to be of very low safety significance since it did not degrade Entergy's ability to terminate a leak path or add reactor coolant inventory when needed, or degrade Entergy's ability to recover RHR once it is was lost. This finding is associated with the cross-cutting area of human performance, in that maintenance personnel did not implement a 480 volt breaker PM procedure correctly.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation MULTIPLE DEFICIENCIES IN SURVEILLANCE PROCEDURES ASSOCIATED WITH ITS CONVERSION The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criterion VI, "Document Control." Inadequate document control resulted in multiple surveillance procedures not meeting the criteria of the Improved Technical Specifications (ITS) surveillance requirements (SRs) or the applicable ITS basis document.
The finding is more than minor since, if left uncorrected, it would become a more significant safety concern potentially impacting multiple SRs of safety-related equipment and equipment important to safety. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality. This finding is considered to be of very low risk significance (Green) since it had not resulted in a loss of safety function or in any inoperable equipment.
Inspection Report# : 2004012(pdf)
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 5 of 7 Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate and timely corrective actions for known deficiencies in the control program(s) and installation of safety related electrical cables and raceways.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergy's failure to promptly take actions to address conditions adverse to quality concerning one example of resolution of Data Verification Transfer Report (DVTR) Items/Operability Assessments; and two examples of configuration control of electrical raceways and cables.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement appropriate design controls for electrical cable and raceway installations.
Green. The team identified three examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for Entergy's failure to implement appropriate design controls for the installation of safety related electrical cables and raceways.
Inspection Report# : 2004009(pdf)
Significance:        Jul 20, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly control cable separation program documents.
Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for Entergy's failure to properly control the cable separation program documents. These documents include some reports never being reviewed, approved, and signed off as well as documents used in part for design specifications and DBD work not entered into the document control program to ensure retrieveability.
Inspection Report# : 2004009(pdf)
Barrier Integrity Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding INEFFECTIVE CAUSAL ANALYSIS ASSOCIATED WITH A ROD CONTROL FAILURE The inspectors identified a Green finding associated with ineffective causal analysis for a rod control system problem which resulted in the unexpected insertion of control rod H-8, and power reductions to less than 75 percent, on February 9 and 10. The inspectors determined that the causal analysis was ineffective since it failed to identify that the current traces taken during troubleshooting were ten to fifteen percent below the expected values, even after short-term action to install the original style regulation cards.
The finding is more than minor since it affected the Barrier Integrity cornerstone objective (fuel cladding). The barrier integrity cornerstone objective provides reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events.
This finding impacted the configuration control attribute since it led to the licensee's inability to maintain the rod alignment criteria prescribed in the Technical Specifications (TS). A Phase 1 SDP screening determined that the inadequate causal analysis and subsequent rod drops were of very low risk significance (Green) since the required actions for rod misalignments prescribed by the TS were performed within the allowed time and in-core flux maps verified that local power limits were met. No violations of NRC requirements were identified. This finding is associated with the cross-cutting area of problem identification and resolution, specifically, an ineffective evaluation of rod control system problems resulted in the unexpected insertion of control rod H-8 and power reductions to less than 75 percent, on February 9 and 10.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS INVOLVING LEAKAGE FROM A CANOPY SEAL WELD ONTO THE REACTOR
 
2Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 6 of 7 VESSEL HEAD IN NOVEMBER 2002 The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for Entergy's failure to properly address a condition adverse to quality involving leakage from a canopy seal weld in November 2002. The ineffective corrective actions for this conoseal leak led to boron accumulation on the reactor vessel head (RVH).
The finding is considered to be more than minor since, if left uncorrected, it could have led to a more significant problem. Specifically, the boric acid, if re-wetted, could have led to accelerated corrosion of the RVH. The finding is of very low significance since the RVH integrity was not affected by this problem. The finding is associated with the cross-cutting area of PI&R related to the ineffective corrective actions for the conoseal leak.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE INSPECTION CRITERIA AND GUIDANCE TO EVALUATORS PRIOR TO THE INSPECTION OF THE REACTOR VESSEL LOWER HEAD PENETRATION NOZZLES The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion IX, "Control of Special Processes," for Entergy's failure to provide adequate inspection criteria and guidance to evaluators prior to the inspection of the reactor vessel lower head penetration nozzles. In particular, Entergy personnel performed visual inspections of the reactor vessel bottom mounted instrumentation annulus area without adequate procedural guidance to define potential problems or indications.
This finding is considered to be more than minor since inspection program deficiencies could allow a degraded component to remain inservice undetected. Specifically, the failure to develop adequate inspection guidance could result in a failure to detect a degraded lower RVH penetration boundary. The finding is of very low significance since the lower RVH integrity was not affected.
Inspection Report# : 2004012(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        Apr 02, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation ENTERGY IP2 DID NOT PROPERLY PACKAGE RADIOACTIVE WASTE FOR DISPOSAL TO CONFORM WITH THE WASTE DISPOSAL FACILITY LICENSE A Green self-revealing non-cited violation of 10 CFR 20.2001 was identified associated with the transfer of waste, by Entergy's Indian Point Energy Center, for disposal, that did not meet Barnwell Low-Level Waste Disposal facility license requirements as required by 10 CFR 30.41.
Specifically, a shipment (0205-12578) of low-level radioactive waste, from the Indian Point Energy Center, was identified on February 11, 2005, at the Barnwell Low-level Waste Disposal Facility, to have loose radioactive waste material inside the shipping cask (and outside of the waste disposal container) contrary to the disposal facility's site operating license (License No. 097, Amendment 47, Condition 61).
This finding is considered to be more than minor because Entergy failed to meet a waste disposal facility license requirement that was reasonably within its ability to foresee, correct, and prevent. This radioactive material control transportation finding was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix D, and determined to be of very low safety significance (Green) because: 1) no external radiation or contamination limits were exceeded; 2) no package breach was involved; 3) no failure to make a notification was involved; and 4) although a low-level burial ground non-conformance was involved, burial ground access was not denied and no 10 CFR 61.55 waste classification issue was involved. In addition, although the finding did involve a certificate of compliance issue; the finding was a minor contents deficiency with low risk significance relative to causing a radioactive release to the public or public or occupational exposure. The small quantity of waste material was contained within the NRC approved shipping cask. Entergy temporarily suspended this type of shipment from the Indian Point Energy Center and placed the issue in the corrective action program.
Inspection Report# : 2005002(pdf)
 
2Q/2005 Inspection Findings - Indian Point 2            Page 7 of 7 Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 24, 2005
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                              Page 1 of 6 Indian Point 2 3Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE CAUSAL ANALYSIS ASSOCIATED WITH STATOR WATER COOLING PRESSURE SWITCH The inspector identified a self-revealing Green finding involving poor causal analysis associated with the main generator stator water cooling (SWC) system. The ineffective causal analysis was associated with the settings of the generator protection trip pressure switch (63-P79). The finding resulted in an automatic reactor trip due to a low inlet pressure condition on the main generator SWC system.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions, and is associated with the equipment performance attribute. Specifically, the finding affects the likelihood of a reactor trip and challenges the critical safety function of auxiliary feedwater (AFW) initiation. The finding is of very low risk significance (Green) since it does not contribute to both the likelihood of a reactor trip and the likelihood of mitigation equipment functions being unavailable. The finding is associated with the cross-cutting area of problem identification and resolution (PI&R) based on the ineffective causal analysis for previously identified deficiencies affecting the SWC system.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation IMPROPER INSTALLATION OF REACTOR COOLANT SYSTEM LOOP FLOW TUBING RESULTING IN REACTOR COOLANT SYSTEM LEAKAGE The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V "Instructions, Procedures and Drawings." Maintenance personnel did not verify that the length of tubing between the RACK 20 bulkhead connection and the existing 21 Reactor Coolant Loop Flow (FT-416) Hi side impulse tubing was sufficient for a proper Swagelok connection pursuant to procedure IP-SMM-MA-108.
The finding is more than minor since it impacts the Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability and challenges critical safety functions, and is associated with the procedural quality attribute. Specifically, the finding affects the likelihood of a reactor coolant system (RCS) leak that upsets plant stability and challenges critical safety functions. This finding is of very low risk significance (Green) since worst case degradation would not result in exceeding the technical specification (TS) limit for identified leakage (10 gpm) and it does not affect the mitigation system's safety functions.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW RCS DRAINDOWN PROCEDURE DUE TO INAPPROPIATE APPROACH The inspector identified a self-revealing Green non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings" associated with a reactor vessel water level control issue during the drain down for the reactor vessel head re-installation on November 11, 2004. Specifically, an inappropriate level reduction rate existed by procedure, such that when communications to field operational personnel were temporarily lost and manual valve manipulations to reduce the rate were delayed, a two foot lower reactor vessel water level resulted.
This finding is more than minor, because it potentially affects the Initiating Events cornerstone objective of limiting the likelihood of events that challenge critical safety functions during shutdown, and is associated with the procedural quality attribute. This finding is considered to be of very low safety significance (Green), because residual heat removal (RHR) shutdown cooling remained operable and gravity re-flood of the reactor without operator action would have limited the consequences of any potential loss of shutdown cooling.
Inspection Report# : 2004012(pdf)
Mitigating Systems
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 2 of 6 Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation INCORRECT SETTING OF RELIEF VALVE SI-855 ABOVE SYSTEM DESIGN PRESSURE AND FAILURE TO SUBMIT REQUIRED CHANGES TO THE SAFETY ANALYSIS REPORT The inspector identified a Green NCV for the licensee's failure to properly implement a design modification involving the Safety Injection pump discharge relief valve, SI-855. This was determined to be a violation of 10CFR50, Appendix B, Part III, Design Control.
The deficiency was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of the SI system to prevent undesirable conditions. The issue was a design deficiency that did not result in loss of function per GL 91-18 (Rev. 1), and was determined to be of very low safety significance since revised calculations demonstrated the system piping remained capable of performing its specified function.
Inspection Report# : 2005004(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post work test resulting in a safety related system exceeding its AOT The inspector identified a Green NCV of 10 CFR 50, App. B, Criterion XI "Test Control" involving an inadequate post work test following maintenance on auxiliary component cooling water discharge check valve 755A. This resulted in the failure to identify a condition which led to one train of the containment recirculation spray system being unavailable for greater than its technical specification (TS) allowed outage time.
The finding is associated with the cross-cutting issue of problem identification and resolution in that the licensee's evaluation for CR IP2-2005-00252 failed to identify the deficiencies in the post maintenance test therefore no corrective actions were written to address this issue until prompted by the inspectors.
This issue is greater than minor because the performance deficiency adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone objective associated with ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP analysis was used to assess the deficiency due to modeling limitations of the Phase 2 SDP tools. The Phase 3 evaluation, performed by a Region I Senior Reactor Analyst, confirmed that this issue was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions associated with training, procedural adequacy and operator knowledge on methods to address degraded grid The inspectors identified a green finding involving inadequate corrective actions associated with training and operator knowledge of plant procedures during degraded grid voltage conditions.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone Objective, and was associated with the cornerstone's procedure quality attribute. The inspectors conducted a Phase 1 SDP screening and determined that the finding was of a very low safety significance, since 138KV system voltage had been maintained greater than the minimum operating voltage throughout the year, therefore implementation of the procedure was not required.
Inspection Report# : 2005003(pdf)
Significance:        May 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE POST-ACCIDENT RECIRCULATION PUMP MOTOR LOADING CONDITIONS USED TO DETERMINE OVERLOAD TRIP SETTINGS FOR 480 VOLT TYPE DB CIRCUIT BREAKERS The team identified a finding where Entergy had used non-conservative post-accident recirculation pump motor loading conditions in an analysis that determined overload trip settings for the associated 480 Volt circuit breakers. This finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion III (Design Control).
This finding is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. This finding is of very low safety significance because it is a design deficiency that did not result in a loss of function.
Inspection Report# : 2005006(pdf)
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 3 of 6 Significance:        May 17, 2005 Identified By: NRC Item Type: AV Apparent Violation FAILURE TO ADEQUATELY EVALUATE AND CORRECT NITROGEN GAS MIGRATION AND ACCUMULATION IN PORTIONS OF THE SAFETY INJECTION SYSTEM An apparent violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Action) and station procedures were identified associated with the failure to evaluate and correct a condition adverse to quality. Specifically, the condition adverse to quality involved the leakage of water from the No. 24 safety injection accumulator past several closed valves, allowing water containing absorbed nitrogen to reach other portions of the safety injection emergency core cooling system (including the common suction supply piping for the safety injection pumps and the 23 safety injection pump casing). As the water moved from a higher to lower system pressure, the nitrogen gas was released from the water, thereby challenging the performance of the safety injection pumps. In addition, Entergys initial evaluation of this condition did not appropriately consider available industry operating experience relative to gas migration into emergency core cooling system piping.
This issue is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The Significance Determination Process (SDP) Phase 1, Phase 2, and Phase 3 were used to determine that this issue represented a finding with preliminarily low to moderate safety significance. The analysis used the NRCs best functionality estimates for the three safety injection pumps over a 17-day period when it was judged that adverse gas accumulation conditions existed. Specifically, the 23 safety injection pump was not functional due to the pump casing being filled with gas. The team concluded that the 21 and 22 pumps, given the accumulated gas in the pump suction piping, would not have functioned 75% of the time (assigned a 75% failure probability) for high flowrate and low discharge pressure conditions in response to a medium break loss of coolant accident; and 25% of the time for low flowrate and high discharge pressure conditions in response to other initiating events. The Phase 1 screening identified that a Phase 2 analysis was needed because the 23 safety injection pump train was not functional for longer than the technical specification allowed outage time of 72 hours. Given the uncertainty in the Phase 2 analysis, a Phase 3 analysis was necessary to improve the accuracy of the result. The Phase 3 analysis for internal and external initiating events, using the above assumptions and licensee risk information, identified an increase in core damage frequency of approximately 1 in 900,000 years of operation (low E-6 per year range); and an increase in large early release frequency of approximately 1 in 3,000,000 years of operation (low E-7 per year range).
Inspection Report# : 2005006(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE INTERIM COMPENSATORY MEASURES FOR FIRE BARRIER IMPAIRMENTS The inspectors identified a Green non-cited violation of license condition 2.K between November 26, 2004 - March 9, 2005, due to inadequate compensatory actions for a degraded 3-hour rated fire barrier (3M Interam) for penetration H20 concurrent with a degraded hose station nearest to the fire barrier H20. Penetration H20 houses electrical cables needed for the Alternate Safe Shutdown System.
The finding is more than minor since, if left uncorrected, the finding would become a more significant safety concern. The finding affects the Mitigating Systems cornerstone, and its objective of ensuring availability, reliability and capability of systems that respond to initiating events, since both deficiencies contributed to plant risk by decreasing the endurance of the fire barrier and affecting the ability to manually (no automatic suppression capability) fight fires in the electrical penetration room. This issue was of very low risk significance (Green) using phase 1 of the Fire Protection SDP, MC 0612 Appendix F because the barrier was judged to afford greater than 20 minutes of fire endurance protection and low combustible loading was found in the fire area. This finding is associated with the cross-cutting area of human performance (personnel) in that fire protection engineering did not document or implement adequate compensatory measures for the degraded fire barrier and inoperable hose station.
Inspection Report# : 2005002(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding FAILURE TO PERIODICALLY VERIFY THE CAPABILITY OF CITY WATER BACKUP COOLING SAFETY FUNCTION The inspectors identified a Green finding associated with a loss of city water to the primary auxiliary building on January 26, 2005.
Specifically, Entergy failed to periodically verify the capability of a backup cooling water supply for the charging pumps, safety injection pumps and the residual heat removal pumps.
The finding is greater than minor since it affected the Mitigating Systems cornerstone objective of availability of backup cooling to safety pumps in response to a loss of all component cooling water and/or loss of service water event. This finding impacted the procedural quality attribute since no periodic verification existed since 2003 to verify the availability of backup cooling water source, city water. In accordance with IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the Region I Senior Reactor Analyst (SRA) performed a Phase 3 analysis and determined that this finding was of very low risk significance (Green). No violations of NRC requirements were identified.
Inspection Report# : 2005002(pdf)
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 4 of 6 Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE PROCEDURES FOR EMERGENCY CORE COOLING SYSTEMS OPERATIONS The inspector identified a Green non-cited violation of TS 5.4.1 associated with Entergy's failure to properly implement procedure 2-COL 10.0, "Locked Safeguards Valves." Residual heat removal recirculation valve AC-1863 was left in the shut position during the restart from IP2 refueling outage No. 16 (2RF16). The valve was not locked open in accordance with 2-COL 10.0 prior to entering Mode 4 due to the sequence of procedures performed at the end of the refueling outage.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of configuration control and adversely affects the capability of systems that respond to initiating events to prevent undesirable consequences. The finding involves the unavailability of a design feature described in the Final Safety Analysis Report (FSAR) that would ensure the capability to continue high-head recirculation after a loss of coolant accident (LOCA) in the event of certain system failures. This finding is of very low safety significance (Green), because the normal flow paths for establishing flow to the safety injection (SI) pump suctions during high-head recirculation remained available for the duration of the period that valve AC-1863 was shut. This finding is associated with the cross-cutting area of human performance, in that, operators did not adequately assess a change in the sequence of procedures performed during the refueling outage.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE MAINTENANCE PROCEDURE RESULTING IN ALL EDG'S BEING DECLARED INOPERABLE DUE TO DEFEATING SBO LOGIC The inspector identified a self-revealing Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings." A maintenance procedure for trip checks associated with the 345KV electrical feeder was inadequate since it did not provide appropriate directions for the test set-up. As a result, technicians unintentionally reset the main generator lock-out relays by using test stabs which defeated the station blackout (SBO) relays associated with the emergency diesel generators (EDGs) starting logic.
The finding is more than minor since it affects the procedure quality attribute of the Mitigating Systems cornerstone and impacts the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance (Green) due to low exposure time, credit for manual actions in the abnormal operating procedures (AOPs) to restore power to the safety-related 480 volt buses and start the required loads to stabilize plant conditions, and the availability of other mitigating equipment (ie. steam driven AFW pump and gas turbines 1 and 2).
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE PREVENTATIVE MAINTENANCE PROCEDURE IMPLEMENTATION RESULTING IN A LOSS OF SAFEGUARDS BUS 6A The inspector identified a self-revealing Green, non-cited violation of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding involved improper maintenance on a 480 volt cross-tie breaker (52/3AT6A). Maintenance personnel did not install the main line contactors for breaker (52/3AT6A) consistent with maintenance procedure BRK-P-003-A, "Westinghouse Model DB-75 Breaker -
Preventative Maintenance."
The finding is greater than minor since it affects the Mitigating Systems cornerstone objective of ensuring the availability of the RHR system and to prevent undesirable consequences such as core damage due to lack of core cooling during plant shutdown. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality (breaker preventative maintenance (PM) procedure). This finding is considered to be of very low safety significance since it did not degrade Entergy's ability to terminate a leak path or add reactor coolant inventory when needed, or degrade Entergy's ability to recover RHR once it is was lost. This finding is associated with the cross-cutting area of human performance, in that maintenance personnel did not implement a 480 volt breaker PM procedure correctly.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation MULTIPLE DEFICIENCIES IN SURVEILLANCE PROCEDURES ASSOCIATED WITH ITS CONVERSION The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criterion VI, "Document Control." Inadequate document control resulted in multiple surveillance procedures not meeting the criteria of the Improved Technical Specifications (ITS) surveillance requirements (SRs) or the applicable ITS basis document.
The finding is more than minor since, if left uncorrected, it would become a more significant safety concern potentially impacting multiple SRs
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 5 of 6 of safety-related equipment and equipment important to safety. The performance finding affects the Mitigating Systems cornerstone attribute of procedural quality. This finding is considered to be of very low risk significance (Green) since it had not resulted in a loss of safety function or in any inoperable equipment.
Inspection Report# : 2004012(pdf)
Barrier Integrity Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding INEFFECTIVE CAUSAL ANALYSIS ASSOCIATED WITH A ROD CONTROL FAILURE The inspectors identified a Green finding associated with ineffective causal analysis for a rod control system problem which resulted in the unexpected insertion of control rod H-8, and power reductions to less than 75 percent, on February 9 and 10. The inspectors determined that the causal analysis was ineffective since it failed to identify that the current traces taken during troubleshooting were ten to fifteen percent below the expected values, even after short-term action to install the original style regulation cards.
The finding is more than minor since it affected the Barrier Integrity cornerstone objective (fuel cladding). The barrier integrity cornerstone objective provides reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events.
This finding impacted the configuration control attribute since it led to the licensee's inability to maintain the rod alignment criteria prescribed in the Technical Specifications (TS). A Phase 1 SDP screening determined that the inadequate causal analysis and subsequent rod drops were of very low risk significance (Green) since the required actions for rod misalignments prescribed by the TS were performed within the allowed time and in-core flux maps verified that local power limits were met. No violations of NRC requirements were identified. This finding is associated with the cross-cutting area of problem identification and resolution, specifically, an ineffective evaluation of rod control system problems resulted in the unexpected insertion of control rod H-8 and power reductions to less than 75 percent, on February 9 and 10.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTIONS INVOLVING LEAKAGE FROM A CANOPY SEAL WELD ONTO THE REACTOR VESSEL HEAD IN NOVEMBER 2002 The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for Entergy's failure to properly address a condition adverse to quality involving leakage from a canopy seal weld in November 2002. The ineffective corrective actions for this conoseal leak led to boron accumulation on the reactor vessel head (RVH).
The finding is considered to be more than minor since, if left uncorrected, it could have led to a more significant problem. Specifically, the boric acid, if re-wetted, could have led to accelerated corrosion of the RVH. The finding is of very low significance since the RVH integrity was not affected by this problem. The finding is associated with the cross-cutting area of PI&R related to the ineffective corrective actions for the conoseal leak.
Inspection Report# : 2004012(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE INSPECTION CRITERIA AND GUIDANCE TO EVALUATORS PRIOR TO THE INSPECTION OF THE REACTOR VESSEL LOWER HEAD PENETRATION NOZZLES The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion IX, "Control of Special Processes," for Entergy's failure to provide adequate inspection criteria and guidance to evaluators prior to the inspection of the reactor vessel lower head penetration nozzles. In particular, Entergy personnel performed visual inspections of the reactor vessel bottom mounted instrumentation annulus area without adequate procedural guidance to define potential problems or indications.
This finding is considered to be more than minor since inspection program deficiencies could allow a degraded component to remain inservice undetected. Specifically, the failure to develop adequate inspection guidance could result in a failure to detect a degraded lower RVH penetration boundary. The finding is of very low significance since the lower RVH integrity was not affected.
Inspection Report# : 2004012(pdf)
 
3Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 6 of 6 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        Apr 02, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation ENTERGY IP2 DID NOT PROPERLY PACKAGE RADIOACTIVE WASTE FOR DISPOSAL TO CONFORM WITH THE WASTE DISPOSAL FACILITY LICENSE A Green self-revealing non-cited violation of 10 CFR 20.2001 was identified associated with the transfer of waste, by Entergy's Indian Point Energy Center, for disposal, that did not meet Barnwell Low-Level Waste Disposal facility license requirements as required by 10 CFR 30.41.
Specifically, a shipment (0205-12578) of low-level radioactive waste, from the Indian Point Energy Center, was identified on February 11, 2005, at the Barnwell Low-level Waste Disposal Facility, to have loose radioactive waste material inside the shipping cask (and outside of the waste disposal container) contrary to the disposal facility's site operating license (License No. 097, Amendment 47, Condition 61).
This finding is considered to be more than minor because Entergy failed to meet a waste disposal facility license requirement that was reasonably within its ability to foresee, correct, and prevent. This radioactive material control transportation finding was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix D, and determined to be of very low safety significance (Green) because: 1) no external radiation or contamination limits were exceeded; 2) no package breach was involved; 3) no failure to make a notification was involved; and 4) although a low-level burial ground non-conformance was involved, burial ground access was not denied and no 10 CFR 61.55 waste classification issue was involved. In addition, although the finding did involve a certificate of compliance issue; the finding was a minor contents deficiency with low risk significance relative to causing a radioactive release to the public or public or occupational exposure. The small quantity of waste material was contained within the NRC approved shipping cask. Entergy temporarily suspended this type of shipment from the Indian Point Energy Center and placed the issue in the corrective action program.
Inspection Report# : 2005002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : November 30, 2005
 
4Q/2005 Inspection Findings - Indian Point 2                                                                                          Page 1 of 6 Indian Point 2 4Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Maintain Design Control of Control Rod Drive Mechanism Fans The NRC identified a Green finding because Entergy did not maintain appropriate design control of the control rod drive mechanism fans. A modification to improve the reliability of these fans was incorrectly implemented, leading to an increased likelihood of a loss of lubrication to the fans' motor bearings. Incorrect implementation of this modification directly resulted in the failure of one of the fans during plant operation.
In response, Entergy entered this issue into their corrective action program.
This finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of the control rod drive mechanism fans, which have a mitigating function in Indian Point 2's emergency operating procedures, was adversely affected. This finding is of very low safety significance because while equipment reliability was degraded, there was no actual loss of safety function. (Section 1R12)
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Control of Work on Safety-Related Components The NRC identified a Green NCV of Technical Specification 5.4.1 because Indian Point's work control process inappropriately allowed maintenance to be conducted on a safety-related component prior to the completion of a modification package or analysis. Indian Point's procedures allowed maintenance to be declared "emergency work" if that work was necessary to avoid a forced shutdown or plant transient.
This allowed the maintenance to be performed prior to the completion of work procedures, a modification package, and the associated engineering analysis. Entergy entered this issue into the corrective action program and took action to revise their work control procedure.
This finding was determined to be more than minor because if left uncorrected it would become a more significant safety concern. Failure to complete an appropriate evaluation prior to work on safety-related equipment could impact the operability of risk-significance components.
This finding is of very low safety significance because the safety-related work performed without appropriate evaluation did not result in the actual loss of safety function of a system and did not impact fire, flooding, seismic, or severe weather initiating events. (Section 1R13)
This finding is associated with the Human Performance cross-cutting area in that the decision to implement the modification to FCV-447 without proper evaluation was based on inappropriate procedural guidance. This deficiency ultimately led to the violation of Indian Point 2's Technical Specifications. (Section 4OA4)
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation INCORRECT SETTING OF RELIEF VALVE SI-855 ABOVE SYSTEM DESIGN PRESSURE AND FAILURE TO SUBMIT REQUIRED CHANGES TO THE SAFETY ANALYSIS REPORT The inspector identified a Green NCV for the licensee's failure to properly implement a design modification involving the Safety Injection pump discharge relief valve, SI-855. This was determined to be a violation of 10CFR50, Appendix B, Part III, Design Control.
The deficiency was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of the SI system to prevent undesirable conditions. The issue was a design deficiency that did not result in loss of function per GL 91-18 (Rev. 1), and was determined to be of very low safety significance since revised calculations demonstrated the
 
4Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 2 of 6 system piping remained capable of performing its specified function.
Inspection Report# : 2005004(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post work test resulting in a safety related system exceeding its AOT The inspector identified a Green NCV of 10 CFR 50, App. B, Criterion XI "Test Control" involving an inadequate post work test following maintenance on auxiliary component cooling water discharge check valve 755A. This resulted in the failure to identify a condition which led to one train of the containment recirculation spray system being unavailable for greater than its technical specification (TS) allowed outage time.
The finding is associated with the cross-cutting issue of problem identification and resolution in that the licensee's evaluation for CR IP2-2005-00252 failed to identify the deficiencies in the post maintenance test therefore no corrective actions were written to address this issue until prompted by the inspectors.
This issue is greater than minor because the performance deficiency adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone objective associated with ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP analysis was used to assess the deficiency due to modeling limitations of the Phase 2 SDP tools. The Phase 3 evaluation, performed by a Region I Senior Reactor Analyst, confirmed that this issue was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions associated with training, procedural adequacy and operator knowledge on methods to address degraded grid The inspectors identified a green finding involving inadequate corrective actions associated with training and operator knowledge of plant procedures during degraded grid voltage conditions.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone Objective, and was associated with the cornerstone's procedure quality attribute. The inspectors conducted a Phase 1 SDP screening and determined that the finding was of a very low safety significance, since 138KV system voltage had been maintained greater than the minimum operating voltage throughout the year, therefore implementation of the procedure was not required.
Inspection Report# : 2005003(pdf)
Significance:        May 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE POST-ACCIDENT RECIRCULATION PUMP MOTOR LOADING CONDITIONS USED TO DETERMINE OVERLOAD TRIP SETTINGS FOR 480 VOLT TYPE DB CIRCUIT BREAKERS The team identified a finding where Entergy had used non-conservative post-accident recirculation pump motor loading conditions in an analysis that determined overload trip settings for the associated 480 Volt circuit breakers. This finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion III (Design Control).
This finding is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. This finding is of very low safety significance because it is a design deficiency that did not result in a loss of function.
Inspection Report# : 2005006(pdf)
Significance:        May 17, 2005 Identified By: NRC Item Type: VIO Violation FAILURE TO ADEQUATELY EVALUATE AND CORRECT NITROGEN GAS MIGRATION AND ACCUMULATION IN PORTIONS OF THE SAFETY INJECTION SYSTEM A violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Action) and station procedures were identified associated with the failure to evaluate and correct a condition adverse to quality. Specifically, the condition adverse to quality involved the leakage of water from the No. 24 safety injection accumulator past several closed valves, allowing water containing absorbed nitrogen to reach other portions of the safety injection emergency core cooling system (including the common suction supply piping for the safety injection pumps and the 23 safety injection pump casing). As the water moved from a higher to lower system pressure, the nitrogen gas was released from the water, thereby challenging the performance of the safety injection pumps. In addition, Entergys initial evaluation of this condition did not appropriately consider available industry operating experience relative to gas migration into emergency core cooling system piping.
 
4Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 3 of 6 This issue is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The Significance Determination Process (SDP) Phase 1, Phase 2, and Phase 3 were used to determine that this issue represented a finding with preliminarily low to moderate safety significance. The analysis used the NRCs best functionality estimates for the three safety injection pumps over a 17-day period when it was judged that adverse gas accumulation conditions existed. Specifically, the 23 safety injection pump was not functional due to the pump casing being filled with gas. The team concluded that the 21 and 22 pumps, given the accumulated gas in the pump suction piping, would not have functioned 75% of the time (assigned a 75% failure probability) for high flowrate and low discharge pressure conditions in response to a medium break loss of coolant accident; and 25% of the time for low flowrate and high discharge pressure conditions in response to other initiating events. The Phase 1 screening identified that a Phase 2 analysis was needed because the 23 safety injection pump train was not functional for longer than the technical specification allowed outage time of 72 hours. Given the uncertainty in the Phase 2 analysis, a Phase 3 analysis was necessary to improve the accuracy of the result. The Phase 3 analysis for internal and external initiating events, using the above assumptions and licensee risk information, identified an increase in core damage frequency of approximately 1 in 900,000 years of operation (low E-6 per year range); and an increase in large early release frequency of approximately 1 in 3,000,000 years of operation (low E-7 per year range).
This deficiency was indicative of cross-cutting weaknesses in the area of problem identification and resolution (evaluation and corrective action).
Inspection Report# : 2005006(pdf)
Inspection Report# : 2005013(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE INTERIM COMPENSATORY MEASURES FOR FIRE BARRIER IMPAIRMENTS The inspectors identified a Green non-cited violation of license condition 2.K between November 26, 2004 - March 9, 2005, due to inadequate compensatory actions for a degraded 3-hour rated fire barrier (3M Interam) for penetration H20 concurrent with a degraded hose station nearest to the fire barrier H20. Penetration H20 houses electrical cables needed for the Alternate Safe Shutdown System.
The finding is more than minor since, if left uncorrected, the finding would become a more significant safety concern. The finding affects the Mitigating Systems cornerstone, and its objective of ensuring availability, reliability and capability of systems that respond to initiating events, since both deficiencies contributed to plant risk by decreasing the endurance of the fire barrier and affecting the ability to manually (no automatic suppression capability) fight fires in the electrical penetration room. This issue was of very low risk significance (Green) using phase 1 of the Fire Protection SDP, MC 0612 Appendix F because the barrier was judged to afford greater than 20 minutes of fire endurance protection and low combustible loading was found in the fire area. This finding is associated with the cross-cutting area of human performance (personnel) in that fire protection engineering did not document or implement adequate compensatory measures for the degraded fire barrier and inoperable hose station.
Inspection Report# : 2005002(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding FAILURE TO PERIODICALLY VERIFY THE CAPABILITY OF CITY WATER BACKUP COOLING SAFETY FUNCTION The inspectors identified a Green finding associated with a loss of city water to the primary auxiliary building on January 26, 2005.
Specifically, Entergy failed to periodically verify the capability of a backup cooling water supply for the charging pumps, safety injection pumps and the residual heat removal pumps.
The finding is greater than minor since it affected the Mitigating Systems cornerstone objective of availability of backup cooling to safety pumps in response to a loss of all component cooling water and/or loss of service water event. This finding impacted the procedural quality attribute since no periodic verification existed since 2003 to verify the availability of backup cooling water source, city water. In accordance with IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the Region I Senior Reactor Analyst (SRA) performed a Phase 3 analysis and determined that this finding was of very low risk significance (Green). No violations of NRC requirements were identified.
Inspection Report# : 2005002(pdf)
Barrier Integrity
 
4Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 4 of 6 Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedural Requirements During Modification of a Safety-Related Valve The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," because Entergy staff did not follow modification procedures when implementing a temporary alteration to FCV-447, the safety-related feedwater flow control valve to the 24 steam generator. Specifically, while implementing a modification to grind material from the valve actuator cap screw heads, maintenance personnel removed more material than allowed by the modification package, thus invalidating the associated structural integrity analysis. The licensee entered this issue into the corrective action program and completed an operability assessment to show that the affected valve remained operable.
This finding is greater than minor because it is associated with the Barrier Integrity cornerstone attribute of Barrier Performance, and affected the cornerstone objective of ensuring the availability and reliability of components used for containment isolation. Improper implementation of this modification could have resulted in the inability of this valve to perform its safety function. This finding is of very low safety significance because while the modification was incorrectly implemented, subsequent analysis showed that the valve remained operable. (Section 1R13)
This finding is associated with the Human Performance cross-cutting area because the failure to maintain design control was the result of a personnel error. (Section 4OA4)
Inspection Report# : 2005005(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding INEFFECTIVE CAUSAL ANALYSIS ASSOCIATED WITH A ROD CONTROL FAILURE The inspectors identified a Green finding associated with ineffective causal analysis for a rod control system problem which resulted in the unexpected insertion of control rod H-8, and power reductions to less than 75 percent, on February 9 and 10. The inspectors determined that the causal analysis was ineffective since it failed to identify that the current traces taken during troubleshooting were ten to fifteen percent below the expected values, even after short-term action to install the original style regulation cards.
The finding is more than minor since it affected the Barrier Integrity cornerstone objective (fuel cladding). The barrier integrity cornerstone objective provides reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events.
This finding impacted the configuration control attribute since it led to the licensee's inability to maintain the rod alignment criteria prescribed in the Technical Specifications (TS). A Phase 1 SDP screening determined that the inadequate causal analysis and subsequent rod drops were of very low risk significance (Green) since the required actions for rod misalignments prescribed by the TS were performed within the allowed time and in-core flux maps verified that local power limits were met. No violations of NRC requirements were identified. This finding is associated with the cross-cutting area of problem identification and resolution, specifically, an ineffective evaluation of rod control system problems resulted in the unexpected insertion of control rod H-8 and power reductions to less than 75 percent, on February 9 and 10.
Inspection Report# : 2005002(pdf)
Emergency Preparedness Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Facilities and Equipment to Determine Threshold for Emergency Action Level A Green non-cited violation associated with emergency planning standard 10 CFR 50.47(b)(4) was identified because there was no instrumentation readily available to operators to provide indication at the service water intake structure to determine if the threshold was met for the declaration of an unusual event. Because this issue is of very low safety significance and has been entered into the corrective action program, it is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy Entergy entered this issue into the corrective action program as CR-IP3-2005-05380 and has taken corrective actions to install a temporary intake level indication until a permanent level indication can be installed.
This finding is greater than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was determined to have very low safety significance (Green) because this deficiency would have precluded the declaration of a UE. This deficiency did not result in the risk significant planning standard function being lost or degraded. (Section 1EP4)
Inspection Report# : 2005005(pdf)
 
4Q/2005 Inspection Findings - Indian Point 2                                                                                            Page 5 of 6 Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Corrective Actions for Frame Relay System Problems The inspectors identified a Green finding for a failure to implement timely corrective actions for multiple frame relay system problems dating back to 2003. Specifically, for issues related to the reliability of the frame relay system, adequate actions to prevent recurrence were not implemented in a timely manner. Entergy's corrective actions in response to the August 2005 frame relay failures resulted in a more thorough assessment of this issue and reasonable actions to prevent recurrence.
This finding was determined to be more than minor because the finding is associated with the EP cornerstone attribute of facilities and equipment (ANS availability). It could affect the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding is not suitable for Significance Determination Process evaluation but has been reviewed by NRC management and is determined to be a finding of very low safety significance. This issue is not greater than Green because of the short periods that the frame relay system was unavailable and because the alert and notification system design included a secondary method (i.e., back-up radio system) to actuate the sirens. This finding was determined to involve a cross-cutting issue in the area of problem resolution.
Inspection Report# : 2005005(pdf)
Significance: SL-IV Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make a 10 CFR 50.72(b)(3)(xiii) Notification A non-cited violation (NCV) of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting a siren system problem that occurred on August 5, 2005. The inspectors noted the short duration of the siren system problem, the fact that the NRC was informally notified, that back-up route alerting was available, and also that the capability to actuate the sirens via a manual siren initiation method was not lost. Subsequent to this event, the licensee implemented corrective actions to formalize the manual siren system actuation method. Notwithstanding these circumstances, a formal notification to the NRC was required because the normal processes for actuation of the sirens were not available and the licensee did not have formal procedures for, and had limited experience with, a potential alternate siren actuation method.
This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRC's ability to carry out its regulatory mission. The inspectors evaluated the severity of this violation using the criteria contained in Supplement I - Reactor Operations and Section VI.A.1 of the NRC's Enforcement Policy and determined that this finding met the criteria for disposition as a non-cited violation.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Public Radiation Safety Significance:        Apr 02, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation ENTERGY IP2 DID NOT PROPERLY PACKAGE RADIOACTIVE WASTE FOR DISPOSAL TO CONFORM WITH THE WASTE DISPOSAL FACILITY LICENSE A Green self-revealing non-cited violation of 10 CFR 20.2001 was identified associated with the transfer of waste, by Entergy's Indian Point Energy Center, for disposal, that did not meet Barnwell Low-Level Waste Disposal facility license requirements as required by 10 CFR 30.41.
Specifically, a shipment (0205-12578) of low-level radioactive waste, from the Indian Point Energy Center, was identified on February 11, 2005, at the Barnwell Low-level Waste Disposal Facility, to have loose radioactive waste material inside the shipping cask (and outside of the waste disposal container) contrary to the disposal facility's site operating license (License No. 097, Amendment 47, Condition 61).
This finding is considered to be more than minor because Entergy failed to meet a waste disposal facility license requirement that was reasonably within its ability to foresee, correct, and prevent. This radioactive material control transportation finding was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix D, and determined to be of very low safety significance (Green) because: 1) no external radiation or contamination limits were exceeded; 2) no package breach was involved; 3) no failure to make a notification was involved; and 4) although a low-level burial ground non-conformance was involved, burial ground access was not denied and no 10 CFR 61.55 waste classification issue was involved. In addition, although the finding did involve a certificate of compliance issue; the finding was a minor contents deficiency with low risk significance relative to causing a radioactive release to the public or public or occupational exposure. The small quantity of waste material was contained within the NRC approved shipping cask. Entergy temporarily suspended this type of shipment
 
4Q/2005 Inspection Findings - Indian Point 2                                              Page 6 of 6 from the Indian Point Energy Center and placed the issue in the corrective action program.
Inspection Report# : 2005002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 03, 2006
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                              Page 1 of 7 Indian Point 2 1Q/2006 Plant Inspection Findings Initiating Events Significance:          Mar 01, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation SCAFFOLDING CONTROL ISSUE RESULTS IN REACTOR TRIP The NRC identified a Green self-revealing NCV of 10 CFR 50.65(a)(4) because Entergy did not adequately assess the risk associated with scaffold construction activities in the cable spreading room. Entergy procedure IP-SMM-WM-100, Work Management Process, requires a risk assessment for activities that increase the risk of a plant transient. No risk assessment was completed for this work as part of the work planning process, and as a result, no risk management actions were developed. During scaffold construction, a contractor inadvertently bumped a switch which resulted in 12 dropped control rods and a subsequent manual reactor trip. Entergy entered this issue into the corrective action program and took immediate actions to improve control of scaffold construction activities.
This finding is greater than minor because it was similar to Example 4.b. of IMC 0612, Appendix E, Examples of Minor Issues, in that the performance deficiency contributed to an actual reactor trip. This finding is of very low safety significance because while it resulted in a reactor trip, it did not also contribute to the unavailability of mitigating systems. The inspectors determined that this finding had a human performance cross-cutting aspect in that Entergy personnel failed to appropriately incorporate risk insights into planning of work activities in close proximity to trip risk components.
Inspection Report# : 2006002(pdf)
Mitigating Systems Significance:          Mar 01, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EFFECTIVELY CONTROL THE PERFORMANCE OF THE ROD POSITION INDICATION SYSTEM The NRC identified a Green NCV of 10 CFR 50.65(a)(2) because Entergy failed to effectively control the performance of the rod position indication system through the use of appropriate preventative maintenance. This resulted in the failure of seven rod bottom lights to illuminate following a reactor trip, creating an additional challenge to plant operators. Entergy entered this issue into their corrective action program and is taking actions to upgrade their surveillance and maintenance procedures relative to the rod position indication system.
The inspectors determined that this finding was greater than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not result in loss of a system or train safety function and did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because Entergy did not thoroughly evaluate multiple rod position indication bistable failures such that the resolution addressed the causes and extent of condition of problems.
Inspection Report# : 2006002(pdf)
Significance:          Feb 22, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR UTILITY TUNNEL DEGRADATION The NRC identified a Green self-revealing NCV of license condition 2.K. because Entergy did not take adequate corrective actions for degraded fire protection piping in the utility tunnel. This issue contributed to failure of a 10 inch high-pressure fire protection line in the tunnel.
Isolation of this leak resulted in loss of high-pressure fire water to three hose stations in the utility tunnel and three fire hydrants on site.
Entergy entered this issue into their corrective action program and is evaluating plans to assess and upgrade the utility tunnel.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern. This finding is of very low safety significance because the areas that lost high-pressure fire water did not contain safety-related or post-fire safe shutdown equipment. The inspectors determined that this finding had a problem identification and resolution cross-cutting aspect because Entergy did not implement
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                            Page 2 of 7 timely and effective corrective actions for safety issues associated with degraded piping in the utility tunnel.
Inspection Report# : 2006002(pdf)
Significance:        Jan 29, 2006 Identified By: NRC Item Type: NCV NonCited Violation DEGRADED RESIDUAL HEAT REMOVAL PUMP FIRE DOOR The NRC identified a Green NCV of license condition 2.K. because Entergy failed to identify a degraded three-hour rated fire door between the 21 and 22 residual heat removal pump cells. The door, which provides a barrier to fire and hot gases between the two cells, was determined to be inoperable due to a 3/8 inch gap between the door and frame along the lower half of the door. Entergy entered this issue into the corrective action program and realigned the door.
This finding is greater than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding is of very low safety significance because the degradation of the fire barrier was low, based on the gap in the door having minimal impact on its performance and reliability. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the condition of the door in the corrective action system.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Maintain Design Control of Control Rod Drive Mechanism Fans The NRC identified a Green finding associated with Entergys failure to maintain appropriate design control of the control rod drive mechanism fans. A design change to improve the reliability of these fans was incorrectly implemented, impacting lubrication of the fans motor bearings and resulting in the early failure of one of the fans during plant operation. Entergy entered this issue into their corrective action program and ordered properly configured fans for installation during the next outage.
This finding is greater than minor because it is associated with the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of the control rod drive mechanism fans, which are required to cool the control rod drive mechanisms during normal operation and are used in the emergency operating procedures to prevent void formation in the reactor head region during natural circulation cool down, was adversely affected. This finding is of very low safety significance because while equipment reliability was degraded, there was no actual loss of system function, and this issue did not result in a plant transient or reactor trip.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Control of Work on Safety-Related Components The NRC identified a Green NCV of Technical Specification 5.4.1 associated with the Indian Point work control process, which inappropriately allowed implementation of work on safety-related components prior to the approval of work procedures, a modification package, and the associated engineering analysis. Specifically, Indian Points work control procedure allowed maintenance to be declared emergency work, which allowed bypassing of the required work review and approval processes, if that work was necessary to avoid a forced shutdown or plant transient. Entergy entered this issue into the corrective action program and took action to revise their work control procedure to modify their definition of emergency work. This finding is associated with the Human Performance cross-cutting area in that the decision to implement a modification in September 2005, without required evaluations, was based on inappropriate procedural guidance.
This finding is greater than minor, because if left uncorrected it would become a more significant safety concern. Failure to complete required evaluations prior to work on safety-related equipment could impact the operability of risk-significant components. On September 27, 2005, Entergy implemented a modification to FCV-447, a safety-related feedwater control valve, using the emergency work provision of the Indian Point work control procedure. This finding is of very low safety significance, because the safety-related work performed without an approved evaluation did not result in the actual loss of safety function of a system and did not impact fire, flooding, seismic, or severe weather initiating events. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                            Page 3 of 7 INCORRECT SETTING OF RELIEF VALVE SI-855 ABOVE SYSTEM DESIGN PRESSURE AND FAILURE TO SUBMIT REQUIRED CHANGES TO THE SAFETY ANALYSIS REPORT The inspector identified a Green NCV for the licensees failure to properly implement a design modification involving the Safety Injection (SI) pump discharge relief valve, SI-855. This was determined to be a violation of 10CFR50 Appendix B, Part III, Design Control.
The deficiency was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of the SI system to prevent undesirable conditions. The issue was a design deficiency that did not result in loss of function per GL 91-18 (rev 1), and was determined to be of very low safety significance (Green) since revised calculations demonstrated the system piping remained capable of performing its specified function.
Inspection Report# : 2005004(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post work test resulting in a safety related system exceeding its AOT The inspector identified a Green NCV of 10 CFR 50, App. B, Criterion XI "Test Control" involving an inadequate post work test following maintenance on auxiliary component cooling water discharge check valve 755A. This resulted in the failure to identify a condition which led to one train of the containment recirculation spray system being unavailable for greater than its technical specification (TS) allowed outage time.
The finding is associated with the cross-cutting issue of problem identification and resolution in that the licensee's evaluation for CR IP2-2005-00252 failed to identify the deficiencies in the post maintenance test therefore no corrective actions were written to address this issue until prompted by the inspectors.
This issue is greater than minor because the performance deficiency adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone objective associated with ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP analysis was used to assess the deficiency due to modeling limitations of the Phase 2 SDP tools. The Phase 3 evaluation, performed by a Region I Senior Reactor Analyst, confirmed that this issue was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions associated with training, procedural adequacy and operator knowledge on methods to address degraded grid The inspectors identified a Green finding involving inadequate corrective actions associated with the adequacy of plant procedures to be utilized during degraded grid voltage conditions and the operators knowledge of these procedures.
This finding is greater than minor because the performance deficiency adversely impacted the Mitigating Systems Cornerstone objective associated with procedure quality. The inspectors conducted a Phase 1 SDP screening anddetermined that the finding was of very low safety significance. The 138KVsystem voltage had been maintained greater than the minimum operating voltage throughout the year and implementation of the procedure was not required, therefore an actual loss of safety function did not exist during the period in question.
Inspection Report# : 2005003(pdf)
Significance:        May 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE POST-ACCIDENT RECIRCULATION PUMP MOTOR LOADING CONDITIONS USED TO DETERMINE OVERLOAD TRIP SETTINGS FOR 480 VOLT TYPE DB CIRCUIT BREAKERS The team identified a finding where Entergy had used non-conservative post-accident recirculation pump motor loading conditions in an analysis that determined overload trip settings for the associated 480 Volt circuit breakers. This finding was determined to be a violation of 10 CFR 50, Appendix B, Criterion III (Design Control).
This finding is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. This finding is of very low safety significance because it is a design deficiency that did not result in a loss of function.
Inspection Report# : 2005006(pdf)
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                            Page 4 of 7 Significance:        May 17, 2005 Identified By: NRC Item Type: VIO Violation FAILURE TO ADEQUATELY EVALUATE AND CORRECT NITROGEN GAS MIGRATION AND ACCUMULATION IN PORTIONS OF THE SAFETY INJECTION SYSTEM A violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Action) and station procedures were identified associated with the failure to evaluate and correct a condition adverse to quality. Specifically, the condition adverse to quality involved the leakage of water from the No. 24 safety injection accumulator past several closed valves, allowing water containing absorbed nitrogen to reach other portions of the safety injection emergency core cooling system (including the common suction supply piping for the safety injection pumps and the 23 safety injection pump casing). As the water moved from a higher to lower system pressure, the nitrogen gas was released from the water, thereby challenging the performance of the safety injection pumps. In addition, Entergys initial evaluation of this condition did not appropriately consider available industry operating experience relative to gas migration into emergency core cooling system piping.
This issue is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The Significance Determination Process (SDP) Phase 1, Phase 2, and Phase 3 were used to determine that this issue represented a finding with preliminarily low to moderate safety significance. The analysis used the NRCs best functionality estimates for the three safety injection pumps over a 17-day period when it was judged that adverse gas accumulation conditions existed. Specifically, the 23 safety injection pump was not functional due to the pump casing being filled with gas. The team concluded that the 21 and 22 pumps, given the accumulated gas in the pump suction piping, would not have functioned 75% of the time (assigned a 75% failure probability) for high flowrate and low discharge pressure conditions in response to a medium break loss of coolant accident; and 25% of the time for low flowrate and high discharge pressure conditions in response to other initiating events. The Phase 1 screening identified that a Phase 2 analysis was needed because the 23 safety injection pump train was not functional for longer than the technical specification allowed outage time of 72 hours. Given the uncertainty in the Phase 2 analysis, a Phase 3 analysis was necessary to improve the accuracy of the result. The Phase 3 analysis for internal and external initiating events, using the above assumptions and licensee risk information, identified an increase in core damage frequency of approximately 1 in 900,000 years of operation (low E-6 per year range); and an increase in large early release frequency of approximately 1 in 3,000,000 years of operation (low E-7 per year range).
This deficiency was indicative of cross-cutting weaknesses in the area of problem identification and resolution (evaluation and corrective action).
Inspection Report# : 2005006(pdf)
Inspection Report# : 2005013(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT ADEQUATE INTERIM COMPENSATORY MEASURES FOR FIRE BARRIER IMPAIRMENTS The inspectors identified a Green non-cited violation of license condition 2.K between November 26, 2004 - March 9, 2005, due to inadequate compensatory actions for a degraded 3-hour rated fire barrier (3M Interam) for penetration H20 concurrent with a degraded hose station nearest to the fire barrier H20. Penetration H20 houses electrical cables needed for the Alternate Safe Shutdown System.
The finding is more than minor since, if left uncorrected, the finding would become a more significant safety concern. The finding affects the Mitigating Systems cornerstone, and its objective of ensuring availability, reliability and capability of systems that respond to initiating events, since both deficiencies contributed to plant risk by decreasing the endurance of the fire barrier and affecting the ability to manually (no automatic suppression capability) fight fires in the electrical penetration room. This issue was of very low risk significance (Green) using phase 1 of the Fire Protection SDP, MC 0612 Appendix F because the barrier was judged to afford greater than 20 minutes of fire endurance protection and low combustible loading was found in the fire area. This finding is associated with the cross-cutting area of human performance (personnel) in that fire protection engineering did not document or implement adequate compensatory measures for the degraded fire barrier and inoperable hose station.
Inspection Report# : 2005002(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding FAILURE TO PERIODICALLY VERIFY THE CAPABILITY OF CITY WATER BACKUP COOLING SAFETY FUNCTION The inspectors identified a Green finding associated with a loss of city water to the primary auxiliary building on January 26, 2005.
Specifically, Entergy failed to periodically verify the capability of a backup cooling water supply for the charging pumps, safety injection pumps and the residual heat removal pumps.
The finding is greater than minor since it affected the Mitigating Systems cornerstone objective of availability of backup cooling to safety
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                            Page 5 of 7 pumps in response to a loss of all component cooling water and/or loss of service water event. This finding impacted the procedural quality attribute since no periodic verification existed since 2003 to verify the availability of backup cooling water source, city water. In accordance with IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the Region I Senior Reactor Analyst (SRA) performed a Phase 3 analysis and determined that this finding was of very low risk significance (Green). No violations of NRC requirements were identified.
Inspection Report# : 2005002(pdf)
Barrier Integrity Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedural Requirements During Modification of a Safety-Related Valve The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures during implementation of a temporary alteration to FCV-447, the safety-related feedwater flow control valve to the 24 steam generator. Specifically, while implementing a modification to grind material from the valve actuator cap screw heads, maintenance personnel removed more material than allowed by the modification package. This error was not identified by the maintenance workers or engineering personnel upon completion of the modification. Entergy entered this issue into the corrective action program and completed an operability assessment to show that FCV-447 remained operable. This finding is associated with the Human Performance cross-cutting area because the failure to follow procedures was the result of a personnel error during implementation of the modification.
This finding is greater than minor because it is associated with the Barrier Integrity cornerstone attribute of Barrier Performance, and affected the cornerstone objective of ensuring the availability and reliability of components used for containment isolation. Improper implementation of this modification could have resulted in the inability of this valve to perform its safety function. This finding is of very low safety significance because while the modification was incorrectly implemented, subsequent analysis showed that the valve remained operable. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:        Apr 02, 2005 Identified By: NRC Item Type: FIN Finding INEFFECTIVE CAUSAL ANALYSIS ASSOCIATED WITH A ROD CONTROL FAILURE The inspectors identified a Green finding associated with ineffective causal analysis for a rod control system problem which resulted in the unexpected insertion of control rod H-8, and power reductions to less than 75 percent, on February 9 and 10. The inspectors determined that the causal analysis was ineffective since it failed to identify that the current traces taken during troubleshooting were ten to fifteen percent below the expected values, even after short-term action to install the original style regulation cards.
The finding is more than minor since it affected the Barrier Integrity cornerstone objective (fuel cladding). The barrier integrity cornerstone objective provides reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events.
This finding impacted the configuration control attribute since it led to the licensee's inability to maintain the rod alignment criteria prescribed in the Technical Specifications (TS). A Phase 1 SDP screening determined that the inadequate causal analysis and subsequent rod drops were of very low risk significance (Green) since the required actions for rod misalignments prescribed by the TS were performed within the allowed time and in-core flux maps verified that local power limits were met. No violations of NRC requirements were identified. This finding is associated with the cross-cutting area of problem identification and resolution, specifically, an ineffective evaluation of rod control system problems resulted in the unexpected insertion of control rod H-8 and power reductions to less than 75 percent, on February 9 and 10.
Inspection Report# : 2005002(pdf)
Emergency Preparedness Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Facilities and Equipment to Determine Threshold for Emergency Action Level A Green NCV associated with emergency planning standard 10 CFR 50.47(b)(4) was identified by the inspectors, because no established means of indication or procedures were readily available for operators to determine if the service water bay level met the threshold for declaration of an Unusual Event (UE) described in EAL 8.4.3. Entergy installed temporary level indication and entered this issue into its
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                          Page 6 of 7 corrective action program for further evaluation and implementation of long term corrective actions This finding is greater than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. Section 4.4 of Manual Chapter 0609, Appendix B, provides examples for use in assessing emergency preparedness related findings. One example of a Green finding states, The EAL classification process would not declare any Alert or Notification of Unusual Event that should be declared. Since the declaration of an UE based on low service water bay level could have been missed or delayed, this finding was considered consistent with the example provided and was therefore determined to be of very low safety significance (Green). Because this issue is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Corrective Actions for Frame Relay System Problems The inspectors identified a Green finding for a failure to implement timely corrective actions for multiple frame relay system problems dating back to 2003. Specifically, for issues related to the reliability of the frame relay system, adequate actions to prevent recurrence were not implemented in a timely manner. Entergys corrective actions in response to the August 2005 frame relay failures resulted in a more thorough assessment of this issue and reasonable actions to prevent recurrence. This finding was associated with the Problem Identification and Resolution cross-cutting area because it was related to Entergys failure to implement timely corrective actions for reliability issues with the frame relay system.
This finding was determined to be more than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment. It affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding is not suitable for Significance Determination Process evaluation but has been reviewed by NRC management and is determined to be a finding of very low safety significance. This issue is not greater than Green, because of the short periods that the frame relay system was unavailable and, because the alert and notification system design included a secondary method (i.e., back-up radio system) to actuate the sirens.
Inspection Report# : 2005005(pdf)
Significance: SL-IV Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make a 10 CFR 50.72(b)(3)(xiii) Notification A Severity Level IV violation of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting a siren system problem that occurred on August 5, 2005. The inspectors noted that the duration of the siren system problem was short, the NRC was informally notified, the process for back-up route alerting was available, and the capability to actuate the sirens via a manual siren initiation method was not lost. Subsequent to this event, Entergy implemented corrective actions to formalize the manual siren system actuation method. Notwithstanding these circumstances, a formal notification to the NRC was required, because the normal processes for actuation of the sirens were not available and Entergy did not have formal procedures for, and had limited experience with, the manual siren initiation method.
This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. Because this finding is of very low safety significance and has been entered into the corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Public Radiation Safety Significance:        Apr 02, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation ENTERGY IP2 DID NOT PROPERLY PACKAGE RADIOACTIVE WASTE FOR DISPOSAL TO CONFORM WITH THE WASTE DISPOSAL FACILITY LICENSE A Green self-revealing non-cited violation of 10 CFR 20.2001 was identified associated with the transfer of waste, by Entergy's Indian Point
 
1Q/2006 Inspection Findings - Indian Point 2                                                                                          Page 7 of 7 Energy Center, for disposal, that did not meet Barnwell Low-Level Waste Disposal facility license requirements as required by 10 CFR 30.41.
Specifically, a shipment (0205-12578) of low-level radioactive waste, from the Indian Point Energy Center, was identified on February 11, 2005, at the Barnwell Low-level Waste Disposal Facility, to have loose radioactive waste material inside the shipping cask (and outside of the waste disposal container) contrary to the disposal facility's site operating license (License No. 097, Amendment 47, Condition 61).
This finding is considered to be more than minor because Entergy failed to meet a waste disposal facility license requirement that was reasonably within its ability to foresee, correct, and prevent. This radioactive material control transportation finding was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix D, and determined to be of very low safety significance (Green) because: 1) no external radiation or contamination limits were exceeded; 2) no package breach was involved; 3) no failure to make a notification was involved; and 4) although a low-level burial ground non-conformance was involved, burial ground access was not denied and no 10 CFR 61.55 waste classification issue was involved. In addition, although the finding did involve a certificate of compliance issue; the finding was a minor contents deficiency with low risk significance relative to causing a radioactive release to the public or public or occupational exposure. The small quantity of waste material was contained within the NRC approved shipping cask. Entergy temporarily suspended this type of shipment from the Indian Point Energy Center and placed the issue in the corrective action program.
Inspection Report# : 2005002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : May 25, 2006
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                Page 1 of 8 Indian Point 2 2Q/2006 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR PLACING STANDBY MAIN LUBE OIL COOLER IN SERVICE A Green self-revealing finding was identified because Entergys procedure for placing the standby main lube oil cooler in service was inadequate. A deficiency in the procedure resulted in a loss of main feedwater, an automatic start of the motor-driven auxiliary feedwater pumps, and a steam generator level transient. This issue was entered into the corrective action program, and the procedural deficiencies were resolved.
The inspectors determined that this finding was associated with the Initiating Events cornerstone; and, it was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 4.b, since the inadequacies in Entergys procedure caused a plant transient. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergys procedures were not complete and accurate, in that, they failed to ensure the standby main lube cooler was properly filled and vented prior to being placed in service.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR PLACING RHR PUMP SUCTION PRESSURE GAUGES IN SERVICE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergys procedures failed to ensure that the 22 residual heat removal (RHR) pump suction pressure gauge was placed in service prior to starting the system in the shutdown cooling mode of operation. This gauge, which is used to identify degrading RHR pump performance when in shutdown cooling, was left isolated after the plant was depressurized. Entergy placed the pressure gauge in service and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, , Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors] and determined that this finding was of very low safety significance because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for placing the RHR system in the shutdown cooling mode of operation was complete and accurate.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PLANT PROCEDURES FOR IMPLEMENTATION OF COMPENSATORY MEASURES The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures were not followed during the installation of compensatory measures to restore operability of the RHR pumps following the identification of service water piping degradation in the primary auxiliary building. The inspectors also identified multiple deficiencies with the installation and implementation of the compensatory measures. In response, Entergy corrected the deficiencies associated with the compensatory measures and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 3.a, in that, the deficiencies identified with Entergys compensatory measures required significant rework to ensure RHR pump operability. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 2, and determined that the finding was of very low significance because the finding did not degrade the equipment, instrumentation, training, or procedures needed for any shutdown safety function. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not follow plant procedures when implementing a temporary alteration required for
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                              Page 2 of 8 the operability of safety-related equipment.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR VENTING THE REACTOR VESSEL HEAD WHILE SHUTDOWN The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures for reactor coolant system venting following depressurization were inadequate. This resulted in the formation of an 850 gallon void in the reactor vessel head while the plant was shutdown and depressurized. Entergy entered this issue into the corrective action program for evaluation.
The inspectors determined that this finding, which was associated with the Initiating Events cornerstone, was more than minor because if it was left uncorrected, it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 3, and determined that a Phase 2 analysis was needed. The Region I Senior Reactor Analyst performed the Phase 2 analysis using IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, and determined that the finding was of very low safety significance based upon the availability of mitigating systems and the low initiating event (loss of inventory) likelihood. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergys shutdown procedures were not complete and accurate, in that, they failed to ensure the reactor vessel head was adequately vented.
Inspection Report# : 2006003 Significance:        Mar 01, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation SCAFFOLDING CONTROL ISSUE RESULTS IN REACTOR TRIP The NRC identified a Green self-revealing NCV of 10 CFR 50.65(a)(4) because Entergy did not adequately assess the risk associated with scaffold construction activities in the cable spreading room. Entergy procedure IP-SMM-WM-100, Work Management Process, requires a risk assessment for activities that increase the risk of a plant transient. No risk assessment was completed for this work as part of the work planning process, and as a result, no risk management actions were developed. During scaffold construction, a contractor inadvertently bumped a switch which resulted in 12 dropped control rods and a subsequent manual reactor trip. Entergy entered this issue into the corrective action program and took immediate actions to improve control of scaffold construction activities.
This finding is greater than minor because it was similar to Example 4.b. of IMC 0612, Appendix E, Examples of Minor Issues, in that the performance deficiency contributed to an actual reactor trip. This finding is of very low safety significance because while it resulted in a reactor trip, it did not also contribute to the unavailability of mitigating systems. The inspectors determined that this finding had a human performance cross-cutting aspect in that Entergy personnel failed to appropriately incorporate risk insights into planning of work activities in close proximity to trip risk components.
Inspection Report# : 2006002(pdf)
Mitigating Systems Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR DEGRADATION OF SERVICE WATER PIPING A Green self-revealing finding was identified because Entergy failed to take adequate corrective actions for a degraded service water pipe in the primary auxiliary building. Degradation of this pipe was identified in 2003, but was not adequately evaluated or repaired. Consequently, in April of 2006, the continued corrosion of this pipe led to a through-wall leak and, if not corrected, would have challenged the operability of the RHR pumps.
Entergy implemented compensatory measures to protect the RHR pumps, repaired the degraded pipe, and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because if it was left uncorrected it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it represented a qualification deficiency that was confirmed not to result in the loss of operability per Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement timely and effective corrective actions for degraded service water piping in the primary auxiliary building.
Inspection Report# : 2006003
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                  Page 3 of 8 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED RESIDUAL HEAT REMOVAL PUMP CELL FIRE DOOR The inspectors identified a Green NCV of license condition 2.K. because Entergy failed to identify a condition adverse to fire protection related to a degraded fire door between the 21 and 22 RHR pump cells. A similar condition with the same door had been previously identified by the NRC in January 2006. Entergy took actions to correct the degraded fire door and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using IMC 0609 Appendix F, Fire Protection Significance Determination Process, and determined that the finding was of very low safety significance because the fire door, which was moderately degraded, provided a minimum of 20 minutes of fire endurance protection; and, the ignition sources and combustible materials in the RHR pump cells were situated in a manner that the degraded fire door would not have been subject to direct flame impingement. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the degraded condition of the door.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE POST-WORK TEST ON 21 EMERGENCY DIESEL GENERATOR The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergys post-maintenance test on the 21 emergency diesel generator (EDG) following a governor replacement in November 2004 was not adequate to ensure it could perform its intended design function. Subsequent testing showed the EDG could not attain its rated load of 2300 kilowatts. Entergy corrected the deficiency with the 21 EDG, performed a post-maintenance test including a run at 2300 kilowatts, and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significant Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not represent a loss of safety function for a train or system as defined in the plant specific risk-informed inspection notebook; and it was not risk significant due to external event initiators.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS THE RISK OF MAINTENANCE ACTIVITIES ON VALVE SI-869A The inspectors identified a Green NCV of 10 CFR Part 50.65(a)(4) because Entergy did not assess the risk associated with maintenance on the discharge containment isolation valve from the 21 containment spray pump, SI-869A. This maintenance resulted in the unavailability of the 21 containment spray train for a period of approximately 90 minutes. Entergy entered this issue into the corrective action program, conducted an extent of condition review, and completed a causal analysis.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to Example 7.e in IMC 0612, Appendix E, Examples of Minor Issues, in that, the licensees risk assessment failed to consider maintenance activities on components that prevent containment failure. The inspectors evaluated the significance of this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, and determined that the finding was of very low safety significance because the calculated risk deficit was not greater than 1 x 10-6. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately incorporate risk insights into planning work activities on SI-869A in accordance with 10 CFR Part 50.65(a)(4) and the Site Management Manual IP-SMM-WM-101, Online Risk Assessment.
Inspection Report# : 2006003 Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEILLANCE TEST PROCEDURE FOR EMERGENCY DIESEL GENERATORS The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant surveillance procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was not adequate to ensure testing at the appropriate power factor limit prescribed by Technical Specifications Surveillance Requirement 3.8.1.10. Entergy entered this issue into the corrective action program and completed an evaluation to assess the operability of all three EDGs.
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                  Page 4 of 8 The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A,  Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not result in the loss of a system or train safety function; and it did not screen as potentially risk-significant due to external events. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was complete and accurate.
Inspection Report# : 2006003 Significance:        Mar 01, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EFFECTIVELY CONTROL THE PERFORMANCE OF THE ROD POSITION INDICATION SYSTEM The NRC identified a Green NCV of 10 CFR 50.65(a)(2) because Entergy failed to effectively control the performance of the rod position indication system through the use of appropriate preventative maintenance. This resulted in the failure of seven rod bottom lights to illuminate following a reactor trip, creating an additional challenge to plant operators. Entergy entered this issue into their corrective action program and is taking actions to upgrade their surveillance and maintenance procedures relative to the rod position indication system.
The inspectors determined that this finding was greater than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not result in loss of a system or train safety function and did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because Entergy did not thoroughly evaluate multiple rod position indication bistable failures such that the resolution addressed the causes and extent of condition of problems.
Inspection Report# : 2006002(pdf)
Significance:        Feb 22, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR UTILITY TUNNEL DEGRADATION The NRC identified a Green self-revealing NCV of license condition 2.K. because Entergy did not take adequate corrective actions for degraded fire protection piping in the utility tunnel. This issue contributed to failure of a 10 inch high-pressure fire protection line in the tunnel. Isolation of this leak resulted in loss of high-pressure fire water to three hose stations in the utility tunnel and three fire hydrants on site. Entergy entered this issue into their corrective action program and is evaluating plans to assess and upgrade the utility tunnel.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern. This finding is of very low safety significance because the areas that lost high-pressure fire water did not contain safety-related or post-fire safe shutdown equipment. The inspectors determined that this finding had a problem identification and resolution cross-cutting aspect because Entergy did not implement timely and effective corrective actions for safety issues associated with degraded piping in the utility tunnel.
Inspection Report# : 2006002(pdf)
Significance:        Jan 29, 2006 Identified By: NRC Item Type: NCV NonCited Violation DEGRADED RESIDUAL HEAT REMOVAL PUMP FIRE DOOR The NRC identified a Green NCV of license condition 2.K. because Entergy failed to identify a degraded three-hour rated fire door between the 21 and 22 residual heat removal pump cells. The door, which provides a barrier to fire and hot gases between the two cells, was determined to be inoperable due to a 3/8 inch gap between the door and frame along the lower half of the door. Entergy entered this issue into the corrective action program and realigned the door.
This finding is greater than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding is of very low safety significance because the degradation of the fire barrier was low, based on the gap in the door having minimal impact on its performance and reliability. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the condition of the door in the corrective action system.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                  Page 5 of 8 Failure to Maintain Design Control of Control Rod Drive Mechanism Fans The NRC identified a Green finding associated with Entergys failure to maintain appropriate design control of the control rod drive mechanism fans. A design change to improve the reliability of these fans was incorrectly implemented, impacting lubrication of the fans motor bearings and resulting in the early failure of one of the fans during plant operation. Entergy entered this issue into their corrective action program and ordered properly configured fans for installation during the next outage.
This finding is greater than minor because it is associated with the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of the control rod drive mechanism fans, which are required to cool the control rod drive mechanisms during normal operation and are used in the emergency operating procedures to prevent void formation in the reactor head region during natural circulation cool down, was adversely affected. This finding is of very low safety significance because while equipment reliability was degraded, there was no actual loss of system function, and this issue did not result in a plant transient or reactor trip.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Control of Work on Safety-Related Components The NRC identified a Green NCV of Technical Specification 5.4.1 associated with the Indian Point work control process, which inappropriately allowed implementation of work on safety-related components prior to the approval of work procedures, a modification package, and the associated engineering analysis. Specifically, Indian Points work control procedure allowed maintenance to be declared emergency work, which allowed bypassing of the required work review and approval processes, if that work was necessary to avoid a forced shutdown or plant transient. Entergy entered this issue into the corrective action program and took action to revise their work control procedure to modify their definition of emergency work. This finding is associated with the Human Performance cross-cutting area in that the decision to implement a modification in September 2005, without required evaluations, was based on inappropriate procedural guidance.
This finding is greater than minor, because if left uncorrected it would become a more significant safety concern. Failure to complete required evaluations prior to work on safety-related equipment could impact the operability of risk-significant components. On September 27, 2005, Entergy implemented a modification to FCV-447, a safety-related feedwater control valve, using the emergency work provision of the Indian Point work control procedure. This finding is of very low safety significance, because the safety-related work performed without an approved evaluation did not result in the actual loss of safety function of a system and did not impact fire, flooding, seismic, or severe weather initiating events. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation INCORRECT SETTING OF RELIEF VALVE SI-855 ABOVE SYSTEM DESIGN PRESSURE AND FAILURE TO SUBMIT REQUIRED CHANGES TO THE SAFETY ANALYSIS REPORT The inspector identified a Green NCV for the licensees failure to properly implement a design modification involving the Safety Injection (SI) pump discharge relief valve, SI-855. This was determined to be a violation of 10CFR50 Appendix B, Part III, Design Control.
The deficiency was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of the SI system to prevent undesirable conditions. The issue was a design deficiency that did not result in loss of function per GL 91-18 (rev 1), and was determined to be of very low safety significance (Green) since revised calculations demonstrated the system piping remained capable of performing its specified function.
Inspection Report# : 2005004(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post work test resulting in a safety related system exceeding its AOT The inspector identified a Green NCV of 10 CFR 50, App. B, Criterion XI "Test Control" involving an inadequate post work test following maintenance on auxiliary component cooling water discharge check valve 755A. This resulted in the failure to identify a condition which led to one train of the containment recirculation spray system being unavailable for greater than its technical specification (TS) allowed outage time.
The finding is associated with the cross-cutting issue of problem identification and resolution in that the licensee's evaluation for CR IP2-2005-00252 failed to identify the deficiencies in the post maintenance test therefore no corrective actions were written to address this issue until prompted by the inspectors.
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                Page 6 of 8 This issue is greater than minor because the performance deficiency adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone objective associated with ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP analysis was used to assess the deficiency due to modeling limitations of the Phase 2 SDP tools. The Phase 3 evaluation, performed by a Region I Senior Reactor Analyst, confirmed that this issue was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Jul 01, 2005 Identified By: NRC Item Type: FIN Finding Inadequate corrective actions associated with training, procedural adequacy and operator knowledge on methods to address degraded grid The inspectors identified a Green finding involving inadequate corrective actions associated with the adequacy of plant procedures to be utilized during degraded grid voltage conditions and the operators knowledge of these procedures.
This finding is greater than minor because the performance deficiency adversely impacted the Mitigating Systems Cornerstone objective associated with procedure quality. The inspectors conducted a Phase 1 SDP screening anddetermined that the finding was of very low safety significance. The 138KVsystem voltage had been maintained greater than the minimum operating voltage throughout the year and implementation of the procedure was not required, therefore an actual loss of safety function did not exist during the period in question.
Inspection Report# : 2005003(pdf)
Barrier Integrity Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedural Requirements During Modification of a Safety-Related Valve The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures during implementation of a temporary alteration to FCV-447, the safety-related feedwater flow control valve to the 24 steam generator.
Specifically, while implementing a modification to grind material from the valve actuator cap screw heads, maintenance personnel removed more material than allowed by the modification package. This error was not identified by the maintenance workers or engineering personnel upon completion of the modification. Entergy entered this issue into the corrective action program and completed an operability assessment to show that FCV-447 remained operable. This finding is associated with the Human Performance cross-cutting area because the failure to follow procedures was the result of a personnel error during implementation of the modification.
This finding is greater than minor because it is associated with the Barrier Integrity cornerstone attribute of Barrier Performance, and affected the cornerstone objective of ensuring the availability and reliability of components used for containment isolation. Improper implementation of this modification could have resulted in the inability of this valve to perform its safety function. This finding is of very low safety significance because while the modification was incorrectly implemented, subsequent analysis showed that the valve remained operable. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Emergency Preparedness Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Facilities and Equipment to Determine Threshold for Emergency Action Level A Green NCV associated with emergency planning standard 10 CFR 50.47(b)(4) was identified by the inspectors, because no established means of indication or procedures were readily available for operators to determine if the service water bay level met the threshold for declaration of an Unusual Event (UE) described in EAL 8.4.3. Entergy installed temporary level indication and entered this issue into its corrective action program for further evaluation and implementation of long term corrective actions This finding is greater than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. Section 4.4 of Manual Chapter 0609, Appendix B, provides examples for use in assessing emergency
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                Page 7 of 8 preparedness related findings. One example of a Green finding states, The EAL classification process would not declare any Alert or Notification of Unusual Event that should be declared. Since the declaration of an UE based on low service water bay level could have been missed or delayed, this finding was considered consistent with the example provided and was therefore determined to be of very low safety significance (Green).
Because this issue is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Corrective Actions for Frame Relay System Problems The inspectors identified a Green finding for a failure to implement timely corrective actions for multiple frame relay system problems dating back to 2003. Specifically, for issues related to the reliability of the frame relay system, adequate actions to prevent recurrence were not implemented in a timely manner. Entergys corrective actions in response to the August 2005 frame relay failures resulted in a more thorough assessment of this issue and reasonable actions to prevent recurrence. This finding was associated with the Problem Identification and Resolution cross-cutting area because it was related to Entergys failure to implement timely corrective actions for reliability issues with the frame relay system.
This finding was determined to be more than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment. It affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding is not suitable for Significance Determination Process evaluation but has been reviewed by NRC management and is determined to be a finding of very low safety significance. This issue is not greater than Green, because of the short periods that the frame relay system was unavailable and, because the alert and notification system design included a secondary method (i.e., back-up radio system) to actuate the sirens.
Inspection Report# : 2005005(pdf)
Significance: SL-IV Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make a 10 CFR 50.72(b)(3)(xiii) Notification A Severity Level IV violation of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting a siren system problem that occurred on August 5, 2005. The inspectors noted that the duration of the siren system problem was short, the NRC was informally notified, the process for back-up route alerting was available, and the capability to actuate the sirens via a manual siren initiation method was not lost. Subsequent to this event, Entergy implemented corrective actions to formalize the manual siren system actuation method. Notwithstanding these circumstances, a formal notification to the NRC was required, because the normal processes for actuation of the sirens were not available and Entergy did not have formal procedures for, and had limited experience with, the manual siren initiation method.
This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. Because this finding is of very low safety significance and has been entered into the corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEY DURING CORE BARREL REPLACEMENT CAUSED UNINTENDED EXPOSURE A Green self-revealing NCV of 10 CFR Part 20.1501, General, was identified because Entergy failed to take adequate radiation surveys during the installation of the core support barrel. As a result, Entergy did not recognize that actual radiological conditions were significantly different than expected, which contributed to unplanned and unintended exposure of a worker. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspector evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that this finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose.
Inspection Report# : 2006003 Significance:        Jun 30, 2006
 
2Q/2006 Inspection Findings - Indian Point 2                                                                                                Page 8 of 8 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PROCEDURAL REQUIREMENTS ASSOCIATED WITH CORE SUPPORT BARREL REPLACEMENT A Green self-revealing NCV of Technical Specification 5.4.1 was identified because Entergy failed to follow procedural requirements during the core support barrel installation activity. As a result, dose rates were significantly higher than expected during the work activity, and a worker received an unplanned and unintended radiation exposure. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergy personnel failed to comply with plant procedures that were required and specified to support reinstallation of the core support barrel.
Inspection Report# : 2006003 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 31, 2006
 
3Q/2006 Inspection Findings - Indian Point 2                                                                          Page 1 of 9 Indian Point 2 3Q/2006 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR PLACING STANDBY MAIN LUBE OIL COOLER IN SERVICE A Green self-revealing finding was identified because Entergys procedure for placing the standby main lube oil cooler in service was inadequate. A deficiency in the procedure resulted in a loss of main feedwater, an automatic start of the motor-driven auxiliary feedwater pumps, and a steam generator level transient. This issue was entered into the corrective action program, and the procedural deficiencies were resolved.
The inspectors determined that this finding was associated with the Initiating Events cornerstone; and, it was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 4.b, since the inadequacies in Entergys procedure caused a plant transient. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergys procedures were not complete and accurate, in that, they failed to ensure the standby main lube cooler was properly filled and vented prior to being placed in service.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR PLACING RHR PUMP SUCTION PRESSURE GAUGES IN SERVICE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergys procedures failed to ensure that the 22 residual heat removal (RHR) pump suction pressure gauge was placed in service prior to starting the system in the shutdown cooling mode of operation. This gauge, which is used to identify degrading RHR pump performance when in shutdown cooling, was left isolated after the plant was depressurized. Entergy placed the pressure gauge in service and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors] and determined that this finding was of very low safety significance because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for placing the RHR system in the shutdown cooling mode of operation was complete and accurate.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2006 Inspection Findings - Indian Point 2                                                                        Page 2 of 9 FAILURE TO FOLLOW PLANT PROCEDURES FOR IMPLEMENTATION OF COMPENSATORY MEASURES The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures were not followed during the installation of compensatory measures to restore operability of the RHR pumps following the identification of service water piping degradation in the primary auxiliary building. The inspectors also identified multiple deficiencies with the installation and implementation of the compensatory measures. In response, Entergy corrected the deficiencies associated with the compensatory measures and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 3.a, in that, the deficiencies identified with Entergys compensatory measures required significant rework to ensure RHR pump operability. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 2, and determined that the finding was of very low significance because the finding did not degrade the equipment, instrumentation, training, or procedures needed for any shutdown safety function. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not follow plant procedures when implementing a temporary alteration required for the operability of safety-related equipment.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR VENTING THE REACTOR VESSEL HEAD WHILE SHUTDOWN The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures for reactor coolant system venting following depressurization were inadequate. This resulted in the formation of an 850 gallon void in the reactor vessel head while the plant was shutdown and depressurized.
Entergy entered this issue into the corrective action program for evaluation.
The inspectors determined that this finding, which was associated with the Initiating Events cornerstone, was more than minor because if it was left uncorrected, it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 3, and determined that a Phase 2 analysis was needed. The Region I Senior Reactor Analyst performed the Phase 2 analysis using IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, and determined that the finding was of very low safety significance based upon the availability of mitigating systems and the low initiating event (loss of inventory) likelihood. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergys shutdown procedures were not complete and accurate, in that, they failed to ensure the reactor vessel head was adequately vented.
Inspection Report# : 2006003(pdf)
Significance:        Mar 01, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation SCAFFOLDING CONTROL ISSUE RESULTS IN REACTOR TRIP The NRC identified a Green self-revealing NCV of 10 CFR 50.65(a)(4) because Entergy did not adequately assess the risk associated with scaffold construction activities in the cable spreading room. Entergy procedure IP-SMM-WM-100, Work Management Process, requires a risk assessment for activities that increase the risk of a plant transient. No risk assessment was completed for this work as part of the work planning process, and as a result, no risk management actions were developed. During scaffold construction, a contractor inadvertently bumped a switch which resulted in 12 dropped control rods and a subsequent manual reactor trip. Entergy entered this issue into the corrective action program and took immediate actions to improve control of scaffold construction activities.
This finding is greater than minor because it was similar to Example 4.b. of IMC 0612, Appendix E, Examples of Minor Issues, in that the performance deficiency contributed to an actual reactor trip. This finding is of very low safety
 
3Q/2006 Inspection Findings - Indian Point 2                                                                            Page 3 of 9 significance because while it resulted in a reactor trip, it did not also contribute to the unavailability of mitigating systems.
The inspectors determined that this finding had a human performance cross-cutting aspect in that Entergy personnel failed to appropriately incorporate risk insights into planning of work activities in close proximity to trip risk components.
Inspection Report# : 2006002(pdf)
Mitigating Systems Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR DEGRADATION OF SERVICE WATER PIPING A Green self-revealing finding was identified because Entergy failed to take adequate corrective actions for a degraded service water pipe in the primary auxiliary building. Degradation of this pipe was identified in 2003, but was not adequately evaluated or repaired. Consequently, in April of 2006, the continued corrosion of this pipe led to a through-wall leak and, if not corrected, would have challenged the operability of the RHR pumps. Entergy implemented compensatory measures to protect the RHR pumps, repaired the degraded pipe, and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because if it was left uncorrected it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it represented a qualification deficiency that was confirmed not to result in the loss of operability per Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement timely and effective corrective actions for degraded service water piping in the primary auxiliary building.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED RESIDUAL HEAT REMOVAL PUMP CELL FIRE DOOR The inspectors identified a Green NCV of license condition 2.K. because Entergy failed to identify a condition adverse to fire protection related to a degraded fire door between the 21 and 22 RHR pump cells. A similar condition with the same door had been previously identified by the NRC in January 2006. Entergy took actions to correct the degraded fire door and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using IMC 0609 Appendix F, Fire Protection Significance Determination Process, and determined that the finding was of very low safety significance because the fire door, which was moderately degraded, provided a minimum of 20 minutes of fire endurance protection; and, the ignition sources and combustible materials in the RHR pump cells were situated in a manner that the degraded fire door would not have been subject to direct flame impingement. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the degraded condition of the door.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2006 Inspection Findings - Indian Point 2                                                                            Page 4 of 9 INADEQUATE POST-WORK TEST ON 21 EMERGENCY DIESEL GENERATOR The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergys post-maintenance test on the 21 emergency diesel generator (EDG) following a governor replacement in November 2004 was not adequate to ensure it could perform its intended design function. Subsequent testing showed the EDG could not attain its rated load of 2300 kilowatts. Entergy corrected the deficiency with the 21 EDG, performed a post-maintenance test including a run at 2300 kilowatts, and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significant Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not represent a loss of safety function for a train or system as defined in the plant specific risk-informed inspection notebook; and it was not risk significant due to external event initiators.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS THE RISK OF MAINTENANCE ACTIVITIES ON VALVE SI-869A The inspectors identified a Green NCV of 10 CFR Part 50.65(a)(4) because Entergy did not assess the risk associated with maintenance on the discharge containment isolation valve from the 21 containment spray pump, SI-869A. This maintenance resulted in the unavailability of the 21 containment spray train for a period of approximately 90 minutes.
Entergy entered this issue into the corrective action program, conducted an extent of condition review, and completed a causal analysis.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to Example 7.e in IMC 0612, Appendix E, Examples of Minor Issues, in that, the licensees risk assessment failed to consider maintenance activities on components that prevent containment failure. The inspectors evaluated the significance of this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, and determined that the finding was of very low safety significance because the calculated risk deficit was not greater than 1 x 10-6. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately incorporate risk insights into planning work activities on SI-869A in accordance with 10 CFR Part 50.65(a)(4) and the Site Management Manual IP-SMM-WM-101, Online Risk Assessment.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEILLANCE TEST PROCEDURE FOR EMERGENCY DIESEL GENERATORS The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant surveillance procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was not adequate to ensure testing at the appropriate power factor limit prescribed by Technical Specifications Surveillance Requirement 3.8.1.10.
Entergy entered this issue into the corrective action program and completed an evaluation to assess the operability of all three EDGs.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A,  Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not result in the loss of a system or train safety function; and it did not screen as potentially risk-significant due to external events. The inspectors also determined that this finding had a cross-cutting
 
3Q/2006 Inspection Findings - Indian Point 2                                                                            Page 5 of 9 aspect in the area of human performance because Entergy did not ensure that procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was complete and accurate.
Inspection Report# : 2006003(pdf)
Significance:        Mar 01, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EFFECTIVELY CONTROL THE PERFORMANCE OF THE ROD POSITION INDICATION SYSTEM The NRC identified a Green NCV of 10 CFR 50.65(a)(2) because Entergy failed to effectively control the performance of the rod position indication system through the use of appropriate preventative maintenance. This resulted in the failure of seven rod bottom lights to illuminate following a reactor trip, creating an additional challenge to plant operators. Entergy entered this issue into their corrective action program and is taking actions to upgrade their surveillance and maintenance procedures relative to the rod position indication system.
The inspectors determined that this finding was greater than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not result in loss of a system or train safety function and did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because Entergy did not thoroughly evaluate multiple rod position indication bistable failures such that the resolution addressed the causes and extent of condition of problems.
Inspection Report# : 2006002(pdf)
Significance:        Feb 22, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR UTILITY TUNNEL DEGRADATION The NRC identified a Green self-revealing NCV of license condition 2.K. because Entergy did not take adequate corrective actions for degraded fire protection piping in the utility tunnel. This issue contributed to failure of a 10 inch high-pressure fire protection line in the tunnel. Isolation of this leak resulted in loss of high-pressure fire water to three hose stations in the utility tunnel and three fire hydrants on site. Entergy entered this issue into their corrective action program and is evaluating plans to assess and upgrade the utility tunnel.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern. This finding is of very low safety significance because the areas that lost high-pressure fire water did not contain safety-related or post-fire safe shutdown equipment. The inspectors determined that this finding had a problem identification and resolution cross-cutting aspect because Entergy did not implement timely and effective corrective actions for safety issues associated with degraded piping in the utility tunnel.
Inspection Report# : 2006002(pdf)
Significance:        Jan 29, 2006 Identified By: NRC Item Type: NCV NonCited Violation DEGRADED RESIDUAL HEAT REMOVAL PUMP FIRE DOOR The NRC identified a Green NCV of license condition 2.K. because Entergy failed to identify a degraded three-hour rated fire door between the 21 and 22 residual heat removal pump cells. The door, which provides a barrier to fire and hot gases between the two cells, was determined to be inoperable due to a 3/8 inch gap between the door and frame along the lower half of the door. Entergy entered this issue into the corrective action program and realigned the door.
This finding is greater than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding is
 
3Q/2006 Inspection Findings - Indian Point 2                                                                        Page 6 of 9 of very low safety significance because the degradation of the fire barrier was low, based on the gap in the door having minimal impact on its performance and reliability. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the condition of the door in the corrective action system.
Inspection Report# : 2006002(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Maintain Design Control of Control Rod Drive Mechanism Fans The NRC identified a Green finding associated with Entergys failure to maintain appropriate design control of the control rod drive mechanism fans. A design change to improve the reliability of these fans was incorrectly implemented, impacting lubrication of the fans motor bearings and resulting in the early failure of one of the fans during plant operation. Entergy entered this issue into their corrective action program and ordered properly configured fans for installation during the next outage.
This finding is greater than minor because it is associated with the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of the control rod drive mechanism fans, which are required to cool the control rod drive mechanisms during normal operation and are used in the emergency operating procedures to prevent void formation in the reactor head region during natural circulation cool down, was adversely affected. This finding is of very low safety significance because while equipment reliability was degraded, there was no actual loss of system function, and this issue did not result in a plant transient or reactor trip.
Inspection Report# : 2005005(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Control of Work on Safety-Related Components The NRC identified a Green NCV of Technical Specification 5.4.1 associated with the Indian Point work control process, which inappropriately allowed implementation of work on safety-related components prior to the approval of work procedures, a modification package, and the associated engineering analysis. Specifically, Indian Points work control procedure allowed maintenance to be declared emergency work, which allowed bypassing of the required work review and approval processes, if that work was necessary to avoid a forced shutdown or plant transient. Entergy entered this issue into the corrective action program and took action to revise their work control procedure to modify their definition of emergency work. This finding is associated with the Human Performance cross-cutting area in that the decision to implement a modification in September 2005, without required evaluations, was based on inappropriate procedural guidance.
This finding is greater than minor, because if left uncorrected it would become a more significant safety concern. Failure to complete required evaluations prior to work on safety-related equipment could impact the operability of risk-significant components. On September 27, 2005, Entergy implemented a modification to FCV-447, a safety-related feedwater control valve, using the emergency work provision of the Indian Point work control procedure. This finding is of very low safety significance, because the safety-related work performed without an approved evaluation did not result in the actual loss of safety function of a system and did not impact fire, flooding, seismic, or severe weather initiating events. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Barrier Integrity
 
3Q/2006 Inspection Findings - Indian Point 2                                                                        Page 7 of 9 Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedural Requirements During Modification of a Safety-Related Valve The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures during implementation of a temporary alteration to FCV-447, the safety-related feedwater flow control valve to the 24 steam generator. Specifically, while implementing a modification to grind material from the valve actuator cap screw heads, maintenance personnel removed more material than allowed by the modification package. This error was not identified by the maintenance workers or engineering personnel upon completion of the modification.
Entergy entered this issue into the corrective action program and completed an operability assessment to show that FCV-447 remained operable. This finding is associated with the Human Performance cross-cutting area because the failure to follow procedures was the result of a personnel error during implementation of the modification.
This finding is greater than minor because it is associated with the Barrier Integrity cornerstone attribute of Barrier Performance, and affected the cornerstone objective of ensuring the availability and reliability of components used for containment isolation. Improper implementation of this modification could have resulted in the inability of this valve to perform its safety function. This finding is of very low safety significance because while the modification was incorrectly implemented, subsequent analysis showed that the valve remained operable. Because this finding is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Emergency Preparedness Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Facilities and Equipment to Determine Threshold for Emergency Action Level A Green NCV associated with emergency planning standard 10 CFR 50.47(b)(4) was identified by the inspectors, because no established means of indication or procedures were readily available for operators to determine if the service water bay level met the threshold for declaration of an Unusual Event (UE) described in EAL 8.4.3. Entergy installed temporary level indication and entered this issue into its corrective action program for further evaluation and implementation of long term corrective actions This finding is greater than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded.
Section 4.4 of Manual Chapter 0609, Appendix B, provides examples for use in assessing emergency preparedness related findings. One example of a Green finding states, The EAL classification process would not declare any Alert or Notification of Unusual Event that should be declared. Since the declaration of an UE based on low service water bay level could have been missed or delayed, this finding was considered consistent with the example provided and was therefore determined to be of very low safety significance (Green). Because this issue is of very low safety significance and has been entered into Entergys corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Corrective Actions for Frame Relay System Problems The inspectors identified a Green finding for a failure to implement timely corrective actions for multiple frame relay system problems dating back to 2003. Specifically, for issues related to the reliability of the frame relay system, adequate
 
3Q/2006 Inspection Findings - Indian Point 2                                                                        Page 8 of 9 actions to prevent recurrence were not implemented in a timely manner. Entergys corrective actions in response to the August 2005 frame relay failures resulted in a more thorough assessment of this issue and reasonable actions to prevent recurrence. This finding was associated with the Problem Identification and Resolution cross-cutting area because it was related to Entergys failure to implement timely corrective actions for reliability issues with the frame relay system.
This finding was determined to be more than minor because it is associated with the Emergency Preparedness cornerstone attribute of Facilities and Equipment. It affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.
This finding is not suitable for Significance Determination Process evaluation but has been reviewed by NRC management and is determined to be a finding of very low safety significance. This issue is not greater than Green, because of the short periods that the frame relay system was unavailable and, because the alert and notification system design included a secondary method (i.e., back-up radio system) to actuate the sirens.
Inspection Report# : 2005005(pdf)
Significance: SL-IV Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make a 10 CFR 50.72(b)(3)(xiii) Notification A Severity Level IV violation of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting a siren system problem that occurred on August 5, 2005. The inspectors noted that the duration of the siren system problem was short, the NRC was informally notified, the process for back-up route alerting was available, and the capability to actuate the sirens via a manual siren initiation method was not lost. Subsequent to this event, Entergy implemented corrective actions to formalize the manual siren system actuation method. Notwithstanding these circumstances, a formal notification to the NRC was required, because the normal processes for actuation of the sirens were not available and Entergy did not have formal procedures for, and had limited experience with, the manual siren initiation method.
This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. Because this finding is of very low safety significance and has been entered into the corrective action program, it is being treated as an NCV.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEY DURING CORE BARREL REPLACEMENT CAUSED UNINTENDED EXPOSURE A Green self-revealing NCV of 10 CFR Part 20.1501, General, was identified because Entergy failed to take adequate radiation surveys during the installation of the core support barrel. As a result, Entergy did not recognize that actual radiological conditions were significantly different than expected, which contributed to unplanned and unintended exposure of a worker. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.
The inspector evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that this finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006
 
3Q/2006 Inspection Findings - Indian Point 2                                                                        Page 9 of 9 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PROCEDURAL REQUIREMENTS ASSOCIATED WITH CORE SUPPORT BARREL REPLACEMENT A Green self-revealing NCV of Technical Specification 5.4.1 was identified because Entergy failed to follow procedural requirements during the core support barrel installation activity. As a result, dose rates were significantly higher than expected during the work activity, and a worker received an unplanned and unintended radiation exposure. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.
The inspectors evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergy personnel failed to comply with plant procedures that were required and specified to support reinstallation of the core support barrel.
Inspection Report# : 2006003(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : December 21, 2006
 
4Q/2006 Inspection Findings - Indian Point 2                                                                        Page 1 of 12 Indian Point 2 4Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE RISK ASSESSMENT FOR 21 MBFP STEAM INLET VALVE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.65(a)(4), because Entergy did not adequately assess and manage the risk of on-line maintenance activities while operating with a degraded steam inlet valve on one of Entergys two main boiler feed pumps (MBFP). Specifically, from November 16 - 21, 2006, the degraded condition of the 21 MBFP increased the likelihood of a reactor trip, but was not assessed or included in the plants on-line risk model. The degraded steam inlet valve allowed the 21 MBFP to operate at full power only if 22 MBFP was also running. On a loss of 22 MBFP, 21 MBFP would not be able to continue feeding the steam generators as designed due to a not fully open high pressure inlet steam valve, which would result in a reactor trip.
Due to the degraded condition of 21 MBFP, Entergy implemented a temporary procedure change that directed operators to trip the reactor if the 22 MBFP stopped while the plant was operating at power. Entergy entered this issue into their corrective action program and properly assessed 21 MBFP risk on December 21, 2006.
The inspectors determined that this finding was more than minor because Entergy failed to consider risk significant SSCs and support systems that were unavailable during the performance of on-line maintenance. Specifically, Entergy failed to assess the increase in online risk from the increased likelihood of a reactor trip due to 21 MBFP degraded condition. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," and determined that this finding was of very low safety significance because the finding resulted in an increase in the incremental core damage probability of less than 1x10-6 (actual increase was 2x10-8).
The inspectors determined that this finding had a Human Performance cross-cutting aspect in that procedural inadequacies existed in the online risk assessment procedure because it did not require degraded equipment which impacted online risk to be evaluated. (Section 1R13)
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE OPERATING PROCEDURES FOR LOSS OF BOTH HEATER DRAIN TANK PUMPS A Green self-revealing finding was identified because Entergy failed to develop adequate procedures for governing the response to a loss of both heater drain tank pumps and to an approaching rod insertion limit (RIL) alarm condition.
Specifically, the procedure governing operator actions during a loss of heater drain tank pumps did not specify for the operators to reset the steam dumps following the rapid downpower. The alarm response procedure for the approaching rod insertion limit condition directed the operators to place the rod control system in manual to stop further automatic inward rod motion. This impacted operators ability to add negative reactivity and control the transient. Entergy entered these procedural deficiencies into their corrective action program and is evaluating the appropriate steps to correct the procedural deficiencies.
The inspectors determined that this finding is greater than minor because it is associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the procedural inadequacies complicated operator actions to a rapid downpower, resulted in a manual reactor trip when the operators determined that they did not have sufficient control of the transient, and could impact other accident sequences requiring negative reactivity
 
4Q/2006 Inspection Findings - Indian Point 2                                                                        Page 2 of 12 addition. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that plant operating procedures were adequate to ensure operators could appropriately respond to a rapid downpower transient.
Inspection Report# : 2006004 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR CALIBRATING THE STEAM DUMP LOSS OF LOAD CONTROLLER A Green self-revealing finding was identified because Entergy failed to develop an accurate procedure for calibration of the steam dump loss of load controller. This resulted in the steam dumps failing to operate properly during a plant transient, complicating operator response, and leading to a manual reactor trip. Following identification of the issue, Entergy entered the issue into the corrective action program, corrected the procedural deficiency, and re-calibrated the controller.
The inspectors determined that this finding is greater than minor because it is associated with the Procedural Quality attribute of the Initiating events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the inadequacy in Entergy's calibration procedure caused the steam dumps to operate improperly during a plant transient and contributed to a reactor trip. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be available. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for calibration of the steam dump loss of load controller was accurate, in that, it specified incorrect settings for the controller.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR PLACING STANDBY MAIN LUBE OIL COOLER IN SERVICE A Green self-revealing finding was identified because Entergys procedure for placing the standby main lube oil cooler in service was inadequate. A deficiency in the procedure resulted in a loss of main feedwater, an automatic start of the motor-driven auxiliary feedwater pumps, and a steam generator level transient. This issue was entered into the corrective action program, and the procedural deficiencies were resolved.
The inspectors determined that this finding was associated with the Initiating Events cornerstone; and, it was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 4.b, since the inadequacies in Entergys procedure caused a plant transient. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergys procedures were not complete and accurate, in that, they failed to ensure the standby main lube cooler was properly filled and vented prior to being placed in service.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR PLACING RHR PUMP SUCTION PRESSURE GAUGES IN SERVICE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50,
 
4Q/2006 Inspection Findings - Indian Point 2                                                                        Page 3 of 12 Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergys procedures failed to ensure that the 22 residual heat removal (RHR) pump suction pressure gauge was placed in service prior to starting the system in the shutdown cooling mode of operation. This gauge, which is used to identify degrading RHR pump performance when in shutdown cooling, was left isolated after the plant was depressurized. Entergy placed the pressure gauge in service and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors] and determined that this finding was of very low safety significance because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for placing the RHR system in the shutdown cooling mode of operation was complete and accurate.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PLANT PROCEDURES FOR IMPLEMENTATION OF COMPENSATORY MEASURES The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures were not followed during the installation of compensatory measures to restore operability of the RHR pumps following the identification of service water piping degradation in the primary auxiliary building. The inspectors also identified multiple deficiencies with the installation and implementation of the compensatory measures. In response, Entergy corrected the deficiencies associated with the compensatory measures and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 3.a, in that, the deficiencies identified with Entergys compensatory measures required significant rework to ensure RHR pump operability. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 2, and determined that the finding was of very low significance because the finding did not degrade the equipment, instrumentation, training, or procedures needed for any shutdown safety function. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not follow plant procedures when implementing a temporary alteration required for the operability of safety-related equipment.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR VENTING THE REACTOR VESSEL HEAD WHILE SHUTDOWN The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures for reactor coolant system venting following depressurization were inadequate. This resulted in the formation of an 850 gallon void in the reactor vessel head while the plant was shutdown and depressurized.
Entergy entered this issue into the corrective action program for evaluation.
The inspectors determined that this finding, which was associated with the Initiating Events cornerstone, was more than minor because if it was left uncorrected, it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 3, and determined that a Phase 2 analysis was needed. The Region I Senior Reactor Analyst performed the Phase 2 analysis using IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, and
 
4Q/2006 Inspection Findings - Indian Point 2                                                                          Page 4 of 12 determined that the finding was of very low safety significance based upon the availability of mitigating systems and the low initiating event (loss of inventory) likelihood. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergys shutdown procedures were not complete and accurate, in that, they failed to ensure the reactor vessel head was adequately vented.
Inspection Report# : 2006003 (pdf)
Significance:        Mar 01, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation SCAFFOLDING CONTROL ISSUE RESULTS IN REACTOR TRIP The NRC identified a Green self-revealing NCV of 10 CFR 50.65(a)(4) because Entergy did not adequately assess the risk associated with scaffold construction activities in the cable spreading room. Entergy procedure IP-SMM-WM-100, Work Management Process, requires a risk assessment for activities that increase the risk of a plant transient. No risk assessment was completed for this work as part of the work planning process, and as a result, no risk management actions were developed. During scaffold construction, a contractor inadvertently bumped a switch which resulted in 12 dropped control rods and a subsequent manual reactor trip. Entergy entered this issue into the corrective action program and took immediate actions to improve control of scaffold construction activities.
This finding is greater than minor because it was similar to Example 4.b. of IMC 0612, Appendix E, Examples of Minor Issues, in that the performance deficiency contributed to an actual reactor trip. This finding is of very low safety significance because while it resulted in a reactor trip, it did not also contribute to the unavailability of mitigating systems.
The inspectors determined that this finding had a human performance cross-cutting aspect in that Entergy personnel failed to appropriately incorporate risk insights into planning of work activities in close proximity to trip risk components.
Inspection Report# : 2006002 (pdf)
Mitigating Systems Significance:        Dec 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO CORRECT A DEGRADED CONDITION WHICH IMPACTED GAS TURBINE #1 RELIABILITY AND AVAILABILITY The inspectors identified a Green finding in that corrective actions were inadequate to repair a deficiency associated with the gas turbine #1 (GT-1) starting diesel. This deficiency was identified following a failure of GT-1 to start on February 7, 2005, and resulted in three subsequent failures. A corrective action was written to correct the deficient condition following the initial failure and was closed on June 22, 2005, with no actions taken based on senior management decision to preempt preventive maintenance activities on the gas turbines due to pending system retirement. Entergy entered this issue into the corrective action program and installed a modification to the coolant system to prevent further trips due to this condition.
The inspectors determined that this finding was more than minor since it was associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it impacted GT-1 reliability in that the deficiency resulted in multiple failures to start on demand after the condition was identified and the action to correct the condition was closed without being implemented. The inspectors conducted a Phase 1 SDP screening and determined that a Phase 2 evaluation was required since the finding represented an actual loss of safety function a non-Technical Specification required train of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours. The Phase 2 approximation yielded a result of very low safety significance (Green).
The inspectors determined this finding had a cross-cutting aspect in the area of Human Performance in that Entergy did not ensure that equipment and resources were available and adequate to assure nuclear safety. Specifically, Entergy did not maintain plant safety through the minimization of long-standing equipment issues and the minimization of maintenance deferrals associated with the gas turbine system. (Section 1R12)
 
4Q/2006 Inspection Findings - Indian Point 2                                                                        Page 5 of 12 Inspection Report# : 2006005 (pdf)
Significance:        Dec 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY A DEGRADED CONDITION OF AN AUXILIARY FEED WATER CHECK VALVE IN THE CORRECTIVE ACTION PROGRAM The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program. Specifically, upon inspection in September 2006, the gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation. Gasket damage was noted in work orders during internal valve inspections of BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in the corrective action program. Consequently, the problem was not evaluated and corrected prior to reassembly of the valve. Entergy entered this issue into the corrective action program, evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
Inspection Report# : 2006006 (pdf)
Significance:        Dec 05, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE EVALUATION OF LEAKING 22 STEAM GENERATOR LOW FLOW BYPASS VALVE FCV-427L A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage into the 22 steam generator. During the Indian Point Unit 2 reactor trip on August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively and resulted in an uncontrolled rise in 22 steam generator level; operator response to isolate feedwater to the steam generator in accordance with emergency operating procedures; and automatic actuation of the feedwater isolation system. The excessive leakage condition into the 22 steam generator was identified on April 4, 2006, prior to Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior to the reactor trip on August 23, 2006. This issue was entered into the corrective action program, and FCV-427L was repaired and retested satisfactorily.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of the finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolutions addressed the causes and extent of condition of the problem.
Inspection Report# : 2006006 (pdf)
 
4Q/2006 Inspection Findings - Indian Point 2                                                                        Page 6 of 12 Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR DEGRADATION OF SERVICE WATER PIPING A Green self-revealing finding was identified because Entergy failed to take adequate corrective actions for a degraded service water pipe in the primary auxiliary building. Degradation of this pipe was identified in 2003, but was not adequately evaluated or repaired. Consequently, in April of 2006, the continued corrosion of this pipe led to a through-wall leak and, if not corrected, would have challenged the operability of the RHR pumps. Entergy implemented compensatory measures to protect the RHR pumps, repaired the degraded pipe, and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because if it was left uncorrected it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it represented a qualification deficiency that was confirmed not to result in the loss of operability per Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement timely and effective corrective actions for degraded service water piping in the primary auxiliary building.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED RESIDUAL HEAT REMOVAL PUMP CELL FIRE DOOR The inspectors identified a Green NCV of license condition 2.K. because Entergy failed to identify a condition adverse to fire protection related to a degraded fire door between the 21 and 22 RHR pump cells. A similar condition with the same door had been previously identified by the NRC in January 2006. Entergy took actions to correct the degraded fire door and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using IMC 0609 Appendix F, Fire Protection Significance Determination Process, and determined that the finding was of very low safety significance because the fire door, which was moderately degraded, provided a minimum of 20 minutes of fire endurance protection; and, the ignition sources and combustible materials in the RHR pump cells were situated in a manner that the degraded fire door would not have been subject to direct flame impingement. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the degraded condition of the door.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE POST-WORK TEST ON 21 EMERGENCY DIESEL GENERATOR The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergys post-maintenance test on the 21 emergency diesel generator (EDG) following a governor replacement in November 2004 was not adequate to ensure it could perform its intended design function. Subsequent testing showed the EDG could not attain its rated load of 2300 kilowatts. Entergy corrected the deficiency with the 21 EDG, performed a post-maintenance test including a run at 2300 kilowatts, and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The
 
4Q/2006 Inspection Findings - Indian Point 2                                                                          Page 7 of 12 inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significant Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not represent a loss of safety function for a train or system as defined in the plant specific risk-informed inspection notebook; and it was not risk significant due to external event initiators.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS THE RISK OF MAINTENANCE ACTIVITIES ON VALVE SI-869A The inspectors identified a Green NCV of 10 CFR Part 50.65(a)(4) because Entergy did not assess the risk associated with maintenance on the discharge containment isolation valve from the 21 containment spray pump, SI-869A. This maintenance resulted in the unavailability of the 21 containment spray train for a period of approximately 90 minutes.
Entergy entered this issue into the corrective action program, conducted an extent of condition review, and completed a causal analysis.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to Example 7.e in IMC 0612, Appendix E, Examples of Minor Issues, in that, the licensees risk assessment failed to consider maintenance activities on components that prevent containment failure. The inspectors evaluated the significance of this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, and determined that the finding was of very low safety significance because the calculated risk deficit was not greater than 1 x 10-6. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately incorporate risk insights into planning work activities on SI-869A in accordance with 10 CFR Part 50.65(a)(4) and the Site Management Manual IP-SMM-WM-101, Online Risk Assessment.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEILLANCE TEST PROCEDURE FOR EMERGENCY DIESEL GENERATORS The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant surveillance procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was not adequate to ensure testing at the appropriate power factor limit prescribed by Technical Specifications Surveillance Requirement 3.8.1.10.
Entergy entered this issue into the corrective action program and completed an evaluation to assess the operability of all three EDGs.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A,  Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not result in the loss of a system or train safety function; and it did not screen as potentially risk-significant due to external events. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was complete and accurate.
Inspection Report# : 2006003 (pdf)
Significance:      Mar 01, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EFFECTIVELY CONTROL THE PERFORMANCE OF THE ROD POSITION INDICATION SYSTEM
 
4Q/2006 Inspection Findings - Indian Point 2                                                                          Page 8 of 12 The NRC identified a Green NCV of 10 CFR 50.65(a)(2) because Entergy failed to effectively control the performance of the rod position indication system through the use of appropriate preventative maintenance. This resulted in the failure of seven rod bottom lights to illuminate following a reactor trip, creating an additional challenge to plant operators. Entergy entered this issue into their corrective action program and is taking actions to upgrade their surveillance and maintenance procedures relative to the rod position indication system.
The inspectors determined that this finding was greater than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not result in loss of a system or train safety function and did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because Entergy did not thoroughly evaluate multiple rod position indication bistable failures such that the resolution addressed the causes and extent of condition of problems.
Inspection Report# : 2006002 (pdf)
Significance:        Feb 22, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR UTILITY TUNNEL DEGRADATION The NRC identified a Green self-revealing NCV of license condition 2.K. because Entergy did not take adequate corrective actions for degraded fire protection piping in the utility tunnel. This issue contributed to failure of a 10 inch high-pressure fire protection line in the tunnel. Isolation of this leak resulted in loss of high-pressure fire water to three hose stations in the utility tunnel and three fire hydrants on site. Entergy entered this issue into their corrective action program and is evaluating plans to assess and upgrade the utility tunnel.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern. This finding is of very low safety significance because the areas that lost high-pressure fire water did not contain safety-related or post-fire safe shutdown equipment. The inspectors determined that this finding had a problem identification and resolution cross-cutting aspect because Entergy did not implement timely and effective corrective actions for safety issues associated with degraded piping in the utility tunnel.
Inspection Report# : 2006002 (pdf)
Significance:        Jan 29, 2006 Identified By: NRC Item Type: NCV NonCited Violation DEGRADED RESIDUAL HEAT REMOVAL PUMP FIRE DOOR The NRC identified a Green NCV of license condition 2.K. because Entergy failed to identify a degraded three-hour rated fire door between the 21 and 22 residual heat removal pump cells. The door, which provides a barrier to fire and hot gases between the two cells, was determined to be inoperable due to a 3/8 inch gap between the door and frame along the lower half of the door. Entergy entered this issue into the corrective action program and realigned the door.
This finding is greater than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding is of very low safety significance because the degradation of the fire barrier was low, based on the gap in the door having minimal impact on its performance and reliability. The inspectors determined that the finding had a problem identification and resolution cross-cutting aspect because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the condition of the door in the corrective action system.
Inspection Report# : 2006002 (pdf)
Barrier Integrity
 
4Q/2006 Inspection Findings - Indian Point 2                                                                      Page 9 of 12 Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONTAINMENT CLOSURE EQUIPMENT The team identified a Severity Level (SL) IV NCV of 10 CFR 50.59, Changes, Tests and Experiments, for failure to obtain a license amendment pursuant to 10 CFR 50.90 prior implementing a change to alter the requirements of a shutdown fission product barrier. The team reviewed Safety Evaluation (SE) 04-0732-MD-00-RE R1, Installation of a Temporary Roll-up Door on the Containment Equipment Hatch, to determine if the conclusion that a licensee amendment was not required was correct. The SE was performed to assess the adequacy of using a roll-up door to meet the requirements of the Technical Specification action statements 3.9.4.A.4 and 3.9.5.B.3. The action statements required that the equipment door or closure plate be properly installed within four hours after a loss of decay heat removal. Specifically, Entergy concluded that the roll-up door was equivalent to the closure plate and, therefore, adequate to close containment as required by the action statement. The team found that the door was not designed to be air-tight, therefore, any radioactive release inside containment would bypass the roll-up door. The team concluded that the roll-up door did not meet the design or licensing basis of the closure plate as described in the Updated Final Safety Analysis Report and previous approved license amendments. Entergy entered the issue into their corrective action program to evaluate and correct.
The team found that Entergy changed the requirements for the shutdown fission product barrier (containment) prior to receiving NRC approval. As a result, traditional enforcement was used to evaluate the issue because the deficiency affected the NRC ability to perform its regulatory function. The severity level of the violation was determined to be SL IV in accordance with example D.5 of Supplement 1 of the NRC Enforcement Policy. Additionally, the issue was determined to be of very low safety significance (Green) based on the low decay heat levels at the time the roll-up door was credited in accordance with IMC 0609 Appendix H Containment Integrity significant determination process. (Section 1R02)
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO SURVEY AND PROVIDE ACCESS TO AN UNPOSTED HIGH RADIATION AREA On August 27, 2006, a self-revealing NCV of 10 CFR 20.1501 with respect to 10 CFR 20.1902(b) was discovered because the licensee failed to conduct a survey to establish the radiological conditions and commensurate postings and controls of an unposted high radiation area that was affected by known changing plant conditions, prior to allowing personnel access to this area. This finding was entered into the licensees corrective action program and training was provided.
The finding is more than minor because it is associated with the occupational radiation safety cornerstone attribute of exposure control and affected the cornerstone objective, because not establishing radiological conditions and commensurate controls after changing plant radiological conditions prior to allowing access to the affected areas can cause increased personnel exposure. The inspector determined that the finding was of very low safety significance (Green) because it did not involve an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. This finding has a cross-cutting aspect in the area of human performance because the licensee did not use a conservative assumption in the decision-making process, in that the Watch RP technician did not question the radiological conditions of the pipe chase area after a change of plant conditions had occurred and did not require a survey of the pipe chase area before authorizing access to personnel. (Section 2OS1)
Inspection Report# : 2006005 (pdf)
 
4Q/2006 Inspection Findings - Indian Point 2                                                                      Page 10 of 12 Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding UNIT 2 CONTAINMENT SUMP STRAINER MODIFICATION COLLECTIVE EXPOSURE OVERRUNS DUE TO INADEQUATE MOD PREPARATION A self-revealing finding was discovered that involved Inadequate modification planning and construction preparations relative to a Unit 2 containment sump strainer modification that resulted in significant unplanned collective exposure (93.7 person-rem compared to a work activity estimate of 10.9 person-rem). The dose overrun was primarily due to inadequate work activity planning. Specifically, the actual job site conditions for installation of the containment sump modification were not adequately evaluated with respect to the radiological impact of increased occupancy in high dose rate work areas.
This resulted in a significant amount of as-found interferences that required removal and reinstallation and differences in as-found dimensions required a significant amount of fit-up problems which required additional in-field high radiation area work. This unplanned additional in-field high radiation work resulted in significant unintended exposure that could have been avoided. This finding was entered into the licensees corrective action program including lessons learned for the Unit 3 containment sump modification.
This finding is more than minor because it resulted in unplanned, unintended collective dose that was greater than 50%
above the intended dose and greater than 5 person-rem due to conditions that were reasonably within the licensees ability to foresee and correct and which should have been prevented. The inspector determined that the finding was of very low safety significance (Green) because: the finding was due to ALARA work control planning and the 3-year rolling average collective dose for Unit 2 was less than 135 person-rem (73 person-rem for 2003-2005). This finding has a cross-cutting aspect in the area of human performance because the licensee did not adequately incorporate job site conditions in the work control planning process. (Section 2OS2)
Inspection Report# : 2006005 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEY DURING CORE BARREL REPLACEMENT CAUSED UNINTENDED EXPOSURE A Green self-revealing NCV of 10 CFR Part 20.1501, General, was identified because Entergy failed to take adequate radiation surveys during the installation of the core support barrel. As a result, Entergy did not recognize that actual radiological conditions were significantly different than expected, which contributed to unplanned and unintended exposure of a worker. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.
The inspector evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that this finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PROCEDURAL REQUIREMENTS ASSOCIATED WITH CORE SUPPORT BARREL REPLACEMENT A Green self-revealing NCV of Technical Specification 5.4.1 was identified because Entergy failed to follow procedural requirements during the core support barrel installation activity. As a result, dose rates were significantly higher than expected during the work activity, and a worker received an unplanned and unintended radiation exposure. Entergy entered this issue into the corrective action program and completed a root cause analysis.
 
4Q/2006 Inspection Findings - Indian Point 2                                                                      Page 11 of 12 The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation.
The inspectors evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergy personnel failed to comply with plant procedures that were required and specified to support reinstallation of the core support barrel.
Inspection Report# : 2006003 (pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Dec 05, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO ENTER SAFETY CULTURE ASSESSMENT RESULTS INTO CORRECTIVE ACTION PROGRAM The NRC inspectors identified a finding when Entergy failed to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for the adverse conditions identified in the 2006 Safety Culture Assessment. Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner. The contractor who performed the independent safety culture assessment presented the site specific results to Entergy management in June 2006. The negative responses and declining trends identified in the assessment constituted adverse conditions that should have been entered into the corrective action program. At the time of the inspection, Entergy had not initiated condition reports for the assessment results. Consequently, the results had not been fully evaluated to understand the causes and identify appropriate actions to address the identified issues. Additionally, organizations identified by the contractor as needing management attention had not developed departmental action plans at the time of the inspection.
Entergy entered this issue into the corrective action program and initiated a learning organization condition report to track development and implementation of action plans to address the assessment results.
The inspectors determined that the finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue, increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance. The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify issues with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner.
Inspection Report# : 2006006 (pdf)
 
4Q/2006 Inspection Findings - Indian Point 2 Page 12 of 12 Last modified : March 01, 2007
 
Indian Point 2 1Q/2007 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE RISK ASSESSMENT FOR 21 MBFP STEAM INLET VALVE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.65(a)(4), because Entergy did not adequately assess and manage the risk of on-line maintenance activities while operating with a degraded steam inlet valve on one of Entergys two main boiler feed pumps (MBFP). Specifically, from November 16 through 21, 2006, the degraded condition of the 21 MBFP increased the likelihood of a reactor trip, but was not assessed or included in the plants on-line risk model. Entergy entered this issue into their corrective action program and properly assessed 21 MBFP risk on November 21, 2006.
The inspectors determined that this finding was more than minor because Entergy failed to consider risk significant structures, systems, components, and support systems that were unavailable during the performance of on-line maintenance. Specifically, Entergy failed to assess the increase in online risk from the increased likelihood of a reactor trip due to the 21 MBFP degraded condition. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," and determined that this finding was of very low safety significance because the finding resulted in an increase in the incremental core damage probability of less than 1x10-6 (actual increase was approximately 2x10-8).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not provide complete and accurate procedures, in that, the online risk assessment procedure did not require degraded equipment that impacted risk to be assessed or managed.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE OPERATING PROCEDURES FOR LOSS OF BOTH HEATER DRAIN TANK PUMPS A Green self-revealing finding was identified because Entergy failed to develop adequate procedures for governing the response to a loss of both heater drain tank pumps and to an approaching rod insertion limit (RIL) alarm condition.
Specifically, the procedure governing operator actions during a loss of heater drain tank pumps did not specify for the operators to reset the steam dumps following the rapid downpower. The alarm response procedure for the approaching rod insertion limit condition directed the operators to place the rod control system in manual to stop further automatic inward rod motion. This impacted operators ability to add negative reactivity and control the transient. Entergy entered these procedural deficiencies into their corrective action program and is evaluating the appropriate steps to correct the procedural deficiencies.
The inspectors determined that this finding is greater than minor because it is associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the procedural inadequacies complicated operator actions to a rapid downpower, resulted in a manual reactor trip when the operators determined that they did not have sufficient control of the transient, and could impact other accident sequences requiring negative reactivity addition. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that plant operating procedures were adequate to ensure operators could appropriately
 
respond to a rapid downpower transient.
Inspection Report# : 2006004 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR CALIBRATING THE STEAM DUMP LOSS OF LOAD CONTROLLER A Green self-revealing finding was identified because Entergy failed to develop an accurate procedure for calibration of the steam dump loss of load controller. This resulted in the steam dumps failing to operate properly during a plant transient, complicating operator response, and leading to a manual reactor trip. Following identification of the issue, Entergy entered the issue into the corrective action program, corrected the procedural deficiency, and re-calibrated the controller.
The inspectors determined that this finding is greater than minor because it is associated with the Procedural Quality attribute of the Initiating events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the inadequacy in Entergy's calibration procedure caused the steam dumps to operate improperly during a plant transient and contributed to a reactor trip. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be available. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for calibration of the steam dump loss of load controller was accurate, in that, it specified incorrect settings for the controller.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR PLACING STANDBY MAIN LUBE OIL COOLER IN SERVICE A Green self-revealing finding was identified because Entergys procedure for placing the standby main lube oil cooler in service was inadequate. A deficiency in the procedure resulted in a loss of main feedwater, an automatic start of the motor-driven auxiliary feedwater pumps, and a steam generator level transient. This issue was entered into the corrective action program, and the procedural deficiencies were resolved.
The inspectors determined that this finding was associated with the Initiating Events cornerstone; and, it was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 4.b, since the inadequacies in Entergys procedure caused a plant transient. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergys procedures were not complete and accurate, in that, they failed to ensure the standby main lube cooler was properly filled and vented prior to being placed in service.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR PLACING RHR PUMP SUCTION PRESSURE GAUGES IN SERVICE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergys procedures failed to ensure that the 22 residual heat removal (RHR) pump suction pressure gauge was placed in service prior to starting the system in the shutdown cooling mode of operation. This gauge, which is used to identify degrading RHR pump performance when in shutdown cooling, was left isolated after the plant was depressurized. Entergy placed the pressure gauge in service and entered the issue into the corrective action program.
 
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors] and determined that this finding was of very low safety significance because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for placing the RHR system in the shutdown cooling mode of operation was complete and accurate.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PLANT PROCEDURES FOR IMPLEMENTATION OF COMPENSATORY MEASURES The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures were not followed during the installation of compensatory measures to restore operability of the RHR pumps following the identification of service water piping degradation in the primary auxiliary building. The inspectors also identified multiple deficiencies with the installation and implementation of the compensatory measures. In response, Entergy corrected the deficiencies associated with the compensatory measures and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 3.a, in that, the deficiencies identified with Entergys compensatory measures required significant rework to ensure RHR pump operability. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 2, and determined that the finding was of very low significance because the finding did not degrade the equipment, instrumentation, training, or procedures needed for any shutdown safety function. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not follow plant procedures when implementing a temporary alteration required for the operability of safety-related equipment.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE FOR VENTING THE REACTOR VESSEL HEAD WHILE SHUTDOWN The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant procedures for reactor coolant system venting following depressurization were inadequate. This resulted in the formation of an 850 gallon void in the reactor vessel head while the plant was shutdown and depressurized.
Entergy entered this issue into the corrective action program for evaluation.
The inspectors determined that this finding, which was associated with the Initiating Events cornerstone, was more than minor because if it was left uncorrected, it would have become a more significant safety concern. The inspectors evaluated the significance of this finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 3, and determined that a Phase 2 analysis was needed. The Region I Senior Reactor Analyst performed the Phase 2 analysis using IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, and determined that the finding was of very low safety significance based upon the availability of mitigating systems and the low initiating event (loss of inventory) likelihood. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergys shutdown procedures were not complete and accurate, in that, they failed to ensure the reactor vessel head was adequately vented.
Inspection Report# : 2006003 (pdf)
 
Mitigating Systems Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL ASSOCIATED WITH VORTEXING AND NET POSITIVE SUCTION HEAD CALCULATIONS The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not ensure adequate suction submergence for the three safety injection (SI) pumps by not properly translating vortex and net positive suction head (NPSH) design parameters into calculations relative to reactor water storage tank (RWST) level. Specifically, Entergy used a non-conservative method to calculate the level required to prevent pump vortexing, and used a non-conservative RWST level value for determining available NPSH for the SI pumps. Entergy entered the issue into their corrective action program and revised the affected calculations.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of the SI pumps, even though the pumps were ultimately shown to be operable. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the significance determination process (SDP), documented in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," because it was a design deficiency that did not result in a loss of SI system operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DIFFERENTIAL PRESSURE VALUE USED FOR MOV 746 AND MOV 747 TONENSURE VALVE CAPABILITY The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not accurately incorporate design parameters into valve thrust calculations for motor operated valve (MOV) 746 and MOV 747. Specifically, Entergy used an incorrect and non-conservative differential pressure in the calculations for MOV 746 and MOV 747, which were developed to verify that the valves could develop sufficient thrust to open under postulated design basis conditions. Additionally, an incorrect equation was used in determining the reduction in motor torque due to degraded voltage conditions. Entergy entered the issue into their corrective action program and revised the affected calculations using the correct information.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of MOV 746 and MOV 747. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it was a design deficiency that did not to result in a loss of MOV 746 and MOV 747 operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR ENVIRONMENTAL EFFECTS TO ENSURE THE AVAILABILITY OF THE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP OPERATION The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not establish adequate design control measures to ensure the availability of the turbine driven auxiliary feedwater pump (TDAFWP) during a postulated loss-of-offsite power (LOOP) event. Under certain LOOP situations, the team determined that the TDAFWP steam supply could be inadvertently isolated
 
because of inadequate calculations and procedures for limiting the AFWP room temperature rise. Specifically, a calculation to determine the auxiliary feedwater pump (AFWP) room temperature rise during a LOOP did not include heat input from the TDAFWP. Further, actions that could limit the rise in AFWP room temperature and prevent the inadvertent isolation of the TDAFW pump (opening an AFWP room roll-up door or promptly restoring forced ventilation) were not included in procedures. Entergy entered this issue into their corrective action program, implemented immediate compensatory actions, and revised AFWP room temperature rise calculations.
The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of safety function of the TDAFWP (single train) for greater than its 72 hour technical specification allowed outage time, based on the teams review and assessment of site ambient temperature data over the last year.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY MONITOR GAS TURBINE SYSTEM PERFORMANCE AS REQUIRED BY THE MAINTENANCE RULE The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50.65(a)
(1), the Maintenance Rule, in that, Entergy failed to monitor the gas turbine (GT) system in a manner that provided reasonable assurance that the system could perform its intended safety function. Specifically, Entergy did not establish appropriate GT reliability goals, and therefore did not take corrective actions, when GT-1 had exceeded these goals for maintenance preventable functions failures (MPFF). In addition, Entergy did not properly classify repeat MPFFs, which resulted in a similar failure to take corrective actions as required. This resulted in additional GT-1 out of service time that would not have happened if appropriate actions had been taken. Entergy entered this issue into their corrective action program and lowered the allowable goal for MPFFs, and revised the GT-1 (a)(1) action plan to improve reliability.
The finding is more than minor because appropriate GT reliability goals were not established commensurate with safety and appropriate corrective actions were not taken when goals were not met. This finding is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, which considered that the additional GT-1 out of service time due to this issue could be as much as three days. The finding has a cross-cutting aspect in the area of human performance because Entergy did not adequately ensure procedures were complete, accurate, and up-to-date.
Specifically, procedure ENN-DC-171, Maintenance Rule Monitoring, did not provide steps to discriminate between the classification of an initial design deficiency and further failures due to the same condition, resulting in mis-classifying several GT functional failures.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT DEGRADED GAS TURBINE 1 RELIABILITY The team identified a finding of very low safety significance involving Entergy procedure, EN-LI-102, Corrective Action Process, in that, Entergy failed to take corrective actions to address degraded GT-1 reliability. This resulted in a two and one half day time period in January 2007 when GT-1 and GT-3 were simultaneously inoperable because, after GT-3 was made inoperable for planned maintenance activities, GT-1 was subsequently found to be inoperable. Specifically, the reliability of GT-1 declined from an average of 75% for 2005 and the first 10 months of 2006, to 50% for the three months from November 2006 to January 2007; however, Entergy did not take actions to correct this degraded reliability. Entergy entered this issue into their corrective action program and developed an action plan to address GT reliability issues.
The issue is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on
 
Phase 1 and Phase 2 of the SDP, assuming that both GT-1 and GT-3 were unavailable for the two and one half days, due to this issue. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not correct degraded reliability of GT-1, resulting in having GT-1 and GT-3 simultaneously inoperable.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE STATION BATTERY CAPACITY TESTING FOR DEGRADATION MONITORING The team identified a finding of very low safety significance (Green) involving a non-cited violation of Technical Specification 3.8.6.6, in that, Entergy did not perform station battery capacity testing in accordance with IEEE Standard 450-1995 (related to battery maintenance and testing). Specifically, Entergy procedurally terminated battery capacity testing at the rated discharge time (four hours), before reaching the minimum voltage, as specified by IEEE Standard 450-1995. This prevented accurate quantitative measurement of capacity degradation and identification of the need to conduct potential accelerated battery testing, as specified by both IEEE Standard 450-1995 and the technical specifications, if battery capacity drops by more than 10% relative to the previous test. Entergy entered the issue into their corrective action program and performed calculations using past test data, which demonstrated that the capacities of station batteries had not degraded more than 10%.
This issue is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of station battery safety function, based upon the teams verification of Entergys calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION FOR HIGH INTER-TIER BATTERY RESISTANCES The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not take effective corrective actions for a condition adverse to quality concerning out-of-tolerance inter-tier resistances on the No. 21 station battery. Specifically, after repeated failures of the No. 21 station battery inter-tier resistance testing, vendor and IEEE Standard 450-1995 recommended corrective actions were not taken to correct the adverse out-of-tolerance resistance trend. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated that the voltage drop due to the as-found resistance of the inter-tier connections was small and did not impact No. 21 battery operability.
This issue is more than minor because if it was left uncorrected, it would have become a more significant safety concern.
Specifically, high resistance connections in a battery that is loaded during accident conditions can cause localized heating and can cause permanent damage to the battery. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 21 station battery safety function, based upon the teams verification of Entergys revised calculations. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not take effective corrective actions to address the adverse trend of out-of-tolerance inter-tier resistances.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS FOR DECREASE IN BATTERY MARGIN The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not promptly identify and correct a condition adverse to quality, with respect to known errors in the No. 23 station battery design calculations. Specifically, Entergy did not recognize at the appropriate time the need to write a condition report, perform an operability determination, or place controls on the use of
 
the No. 23 battery design calculations when errors were discovered in the No. 23 battery design calculations that significantly lowered the battery capacity margin. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated No. 23 station battery operability through the next refueling outage, based on the calculated margin and conservatisms available.
This issue is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 23 station battery safety function, based upon the teams verification of Entergys revised calculations.
The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy failed to promptly identify the decrease in margin found in the No. 23 battery design calculations of record.
Inspection Report# : 2007007 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO CORRECT A DEGRADED CONDITION WHICH IMPACTED GAS TURBINE #1 RELIABILITY AND AVAILABILITY The inspectors identified a Green finding, in that, Entergy's corrective actions were inadequate to resolve a deficiency associated with the gas turbine 1 (GT-1) starting diesel. This deficiency was identified following a failure of GT-1 to start on February 7, 2005, and resulted in three subsequent failures. A corrective action was written to correct the deficient condition following the initial failure and was closed on June 22, 2005, with no actions taken based on a senior management decision to cancel preventive maintenance activities on the gas turbines due to pending system retirement.
Entergy entered this issue into their corrective action program and installed a modification to the coolant system to prevent further trips due to this condition.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it impacted GT-1 reliability, in that, the deficiency resulted in multiple failures to start on demand after the condition was identified and the action to correct the condition was closed without being implemented. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that a Phase 2 evaluation was required because the finding represented an actual loss of safety function of a non-Technical Specification required train of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours. The inspectors used the Risk-Informed Inspection Notebook for Indian Point Nuclear Generating Unit 2, to conduct the Phase 2 evaluation. The inspectors determined that 65 hours of unavailability were caused by the additional failures of GT-1 due to the starting diesel coolant system deficiency. The inspectors conservatively equated this cumulative unavailability time to the total exposure time and used an initiating events likelihood of less than three days. The Phase 2 approximation yielded a result of very low safety significance (Green).
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available and adequate to assure reliable operation of GT-1. Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the gas turbine system.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY A DEGRADED CONDITION OF AN AUXILIARY FEED WATER CHECK VALVE IN THE CORRECTIVE ACTION PROGRAM The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program. Specifically, upon inspection in September 2006, the
 
gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation. Gasket damage was noted in work orders during internal valve inspections of BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in the corrective action program. Consequently, the problem was not evaluated and corrected prior to reassembly of the valve. Entergy entered this issue into the corrective action program, evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
Inspection Report# : 2006006 (pdf)
Significance:        Dec 05, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE EVALUATION OF LEAKING 22 STEAM GENERATOR LOW FLOW BYPASS VALVE FCV-427L A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage into the 22 steam generator. During the Indian Point Unit 2 reactor trip on August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively and resulted in an uncontrolled rise in 22 steam generator level; operator response to isolate feedwater to the steam generator in accordance with emergency operating procedures; and automatic actuation of the feedwater isolation system. The excessive leakage condition into the 22 steam generator was identified on April 4, 2006, prior to Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior to the reactor trip on August 23, 2006. This issue was entered into the corrective action program, and FCV-427L was repaired and retested satisfactorily.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of the finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolutions addressed the causes and extent of condition of the problem.
Inspection Report# : 2006006 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR DEGRADATION OF SERVICE WATER PIPING A Green self-revealing finding was identified because Entergy failed to take adequate corrective actions for a degraded service water pipe in the primary auxiliary building. Degradation of this pipe was identified in 2003, but was not adequately evaluated or repaired. Consequently, in April of 2006, the continued corrosion of this pipe led to a through-wall leak and, if not corrected, would have challenged the operability of the RHR pumps. Entergy implemented compensatory measures to protect the RHR pumps, repaired the degraded pipe, and entered the issue into the corrective action program.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because if it was left uncorrected it would have become a more significant safety concern. The inspectors evaluated
 
the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that the finding was of very low safety significance because it represented a qualification deficiency that was confirmed not to result in the loss of operability per Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement timely and effective corrective actions for degraded service water piping in the primary auxiliary building.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED RESIDUAL HEAT REMOVAL PUMP CELL FIRE DOOR The inspectors identified a Green NCV of license condition 2.K. because Entergy failed to identify a condition adverse to fire protection related to a degraded fire door between the 21 and 22 RHR pump cells. A similar condition with the same door had been previously identified by the NRC in January 2006. Entergy took actions to correct the degraded fire door and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using IMC 0609 Appendix F, Fire Protection Significance Determination Process, and determined that the finding was of very low safety significance because the fire door, which was moderately degraded, provided a minimum of 20 minutes of fire endurance protection; and, the ignition sources and combustible materials in the RHR pump cells were situated in a manner that the degraded fire door would not have been subject to direct flame impingement. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because operators who routinely traverse through the degraded fire door during performance of their rounds had not identified the degraded condition of the door.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE POST-WORK TEST ON 21 EMERGENCY DIESEL GENERATOR The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergys post-maintenance test on the 21 emergency diesel generator (EDG) following a governor replacement in November 2004 was not adequate to ensure it could perform its intended design function. Subsequent testing showed the EDG could not attain its rated load of 2300 kilowatts. Entergy corrected the deficiency with the 21 EDG, performed a post-maintenance test including a run at 2300 kilowatts, and entered the issue into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significant Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not represent a loss of safety function for a train or system as defined in the plant specific risk-informed inspection notebook; and it was not risk significant due to external event initiators.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS THE RISK OF MAINTENANCE ACTIVITIES ON VALVE SI-869A The inspectors identified a Green NCV of 10 CFR Part 50.65(a)(4) because Entergy did not assess the risk associated with maintenance on the discharge containment isolation valve from the 21 containment spray pump, SI-869A. This
 
maintenance resulted in the unavailability of the 21 containment spray train for a period of approximately 90 minutes.
Entergy entered this issue into the corrective action program, conducted an extent of condition review, and completed a causal analysis.
The inspectors determined that this finding, which was associated with the Mitigating Systems cornerstone, was more than minor because it was similar to Example 7.e in IMC 0612, Appendix E, Examples of Minor Issues, in that, the licensees risk assessment failed to consider maintenance activities on components that prevent containment failure. The inspectors evaluated the significance of this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, and determined that the finding was of very low safety significance because the calculated risk deficit was not greater than 1 x 10-6. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately incorporate risk insights into planning work activities on SI-869A in accordance with 10 CFR Part 50.65(a)(4) and the Site Management Manual IP-SMM-WM-101, Online Risk Assessment.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEILLANCE TEST PROCEDURE FOR EMERGENCY DIESEL GENERATORS The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because plant surveillance procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was not adequate to ensure testing at the appropriate power factor limit prescribed by Technical Specifications Surveillance Requirement 3.8.1.10.
Entergy entered this issue into the corrective action program and completed an evaluation to assess the operability of all three EDGs.
The inspectors determined that this finding was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A,  Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that this finding was of very low safety significance because it was not a qualification deficiency; it did not result in the loss of a system or train safety function; and it did not screen as potentially risk-significant due to external events. The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that procedure 2-PT-R84B, 22 EDG 8 Hour Load Run, was complete and accurate.
Inspection Report# : 2006003 (pdf)
Barrier Integrity Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONTAINMENT CLOSURE EQUIPMENT The inspectors identified a Severity Level IV NCV of 10 CFR 50.59, Changes, Tests and Experiments, for failure to obtain a license amendment pursuant to 10 CFR 50.90 prior implementing a change to alter the requirements of a shutdown fission product barrier. The inspectors reviewed Safety Evaluation 04-0732-MD-00-RE R1, Installation of a Temporary Roll-up Door on the Containment Equipment Hatch, to determine if the conclusion that a licensee amendment was not required was correct. Entergy concluded that the roll-up door was equivalent to the closure plate and, therefore, adequate to close containment as required by the action statement. The inspectors found that the door was not designed to be air-tight; therefore, any radioactive release inside containment would bypass the roll-up door. The inspectors concluded that the roll-up door did not meet the design or licensing basis of the closure plate as described in the Updated Final Safety Analysis Report (UFSAR) and previously approved license amendments. Consequently, Entergy incorrectly concluded that a license amendment pursuant to 50.90 was not required prior to implementing the change. Entergy entered the issue into their corrective action program to evaluate and correct.
 
The inspectors determined that Entergy changed the requirements for the shutdown fission product barrier (containment) prior to receiving NRC approval. As a result, traditional enforcement was used to evaluate the issue because the deficiency affected the NRC's ability to perform its regulatory function. The severity level of the violation was determined to be Severity Level IV in accordance with example D.5 of Supplement 1 of the NRC Enforcement Policy. Additionally, the issue was determined to be of very low safety significance (Green) based on the low decay heat levels at the time the roll-up door was credited in accordance with the significance determination process described in Inspection Manual Chapter (IMC) 0609 Appendix H, Containment Integrity.
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO SURVEY AND PROVIDE ACCESS TO AN UNPOSTED HIGH RADIATION AREA A Green, self-revealing NCV of 10 CFR 20.1501 with respect to 10 CFR 20.1902(b) was identified, in that, Entergy failed to survey radiological condition changes after a plant manipulation that was likely to cause a change in radiological conditions, and this led to the failure to post a plant area as a high radiation area. As a result, two workers were allowed access to an unsurveyed and unposted high radiation area.
The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of exposure control and affected the cornerstone objective, because not establishing radiological conditions and commensurate controls after changing plant radiological conditions prior to allowing access to the affected areas can cause increased personnel exposure. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that it was of very low safety significance (Green) because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. This issue was entered into Entergy's corrective action program and training was provided to the radiation protection staff.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not use a conservative assumption in the decision-making process, in that, the watch radiation protection technician did not question the radiological conditions of the pipe chase area after a change of plant conditions had occurred and did not require a survey of the pipe chase area before authorizing access to personnel.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding UNIT 2 CONTAINMENT SUMP STRAINER MODIFICATION COLLECTIVE EXPOSURE OVERRUNS DUE TO INADEQUATE MOD PREPARATION A self-revealing finding was identified that involved inadequate modification planning and construction preparations relative to a Unit 2 containment sump strainer modification that resulted in significant unplanned collective exposure (93.7 person-rem compared to a work activity estimate of 10.9 person-rem). Specifically, the actual job site conditions for installation of the containment sump modification were not adequately evaluated with respect to the radiological impact of increased occupancy in high dose rate work areas. This unplanned additional in-field high radiation work resulted in significant unintended exposure that could have been avoided. This issue was entered into Entergy's corrective action program so that lessons learned could be incorporated into the Unit 3 containment sump modification.
The inspectors determined that this finding was more than minor because it was similar to examples 6.a and 6.b of IMC
 
0612, Appendix E, "Examples of Minor Issues," in that, the issue involved actual collective exposure greater than 5 person-rem and was greater than 50 percent above the estimated or intended exposure; and the majority of the dose overrun was due to activities within Entergy's control. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance (Green) because it involved an ALARA planning issue, and the 3-year rolling average collective dose for Unit 2 was less than 135 person-rem (73 person-rem average annual exposure for 2003 through 2005).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not adequately incorporate job site conditions in the work control planning process.
Inspection Report# : 2006005 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE SURVEY DURING CORE BARREL REPLACEMENT CAUSED UNINTENDED EXPOSURE A Green self-revealing NCV of 10 CFR Part 20.1501, General, was identified because Entergy failed to take adequate radiation surveys during the installation of the core support barrel. As a result, Entergy did not recognize that actual radiological conditions were significantly different than expected, which contributed to unplanned and unintended exposure of a worker. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspector evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that this finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PROCEDURAL REQUIREMENTS ASSOCIATED WITH CORE SUPPORT BARREL REPLACEMENT A Green self-revealing NCV of Technical Specification 5.4.1 was identified because Entergy failed to follow procedural requirements during the core support barrel installation activity. As a result, dose rates were significantly higher than expected during the work activity, and a worker received an unplanned and unintended radiation exposure. Entergy entered this issue into the corrective action program and completed a root cause analysis.
The inspectors determined that this finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone; and, it affected the cornerstone objective of ensuring adequate protection of workers from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors evaluated the significance of this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance because it did not involve: (1) as low as reasonable achievable planning or work controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance because Entergy personnel failed to comply with plant procedures that were required and specified to support reinstallation of the core support barrel.
Inspection Report# : 2006003 (pdf)
Public Radiation Safety
 
Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Dec 05, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO ENTER SAFETY CULTURE ASSESSMENT RESULTS INTO CORRECTIVE ACTION PROGRAM The NRC inspectors identified a finding when Entergy failed to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for the adverse conditions identified in the 2006 Safety Culture Assessment. Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner. The contractor who performed the independent safety culture assessment presented the site specific results to Entergy management in June 2006. The negative responses and declining trends identified in the assessment constituted adverse conditions that should have been entered into the corrective action program. At the time of the inspection, Entergy had not initiated condition reports for the assessment results. Consequently, the results had not been fully evaluated to understand the causes and identify appropriate actions to address the identified issues. Additionally, organizations identified by the contractor as needing management attention had not developed departmental action plans at the time of the inspection.
Entergy entered this issue into the corrective action program and initiated a learning organization condition report to track development and implementation of action plans to address the assessment results.
The inspectors determined that the finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue, increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance. The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify issues with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner.
Inspection Report# : 2006006 (pdf)
Last modified : June 01, 2007
 
Indian Point 2 2Q/2007 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INCORPORATE DESIGN BASIS INFORMATION INTO PROCEDURES TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not appropriately incorporate design requirements into an operating procedure used to establish adequate component cooling water (CCW) flow to the reactor coolant pump (RCP) thermal barriers.
Specifically, the flow specification in the CCW operating procedure did not incorporate the calculated design flow requirements to bound allowable CCW temperature limits. Entergy entered this issue into their corrective action program and will be evaluating the flow requirements specified in procedure 2-SOP-4.1.2, Component Cooling Water System Operation, to ensure that they bound the allowed plant operating limits.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergy did not incorporate design flow requirements necessary to assure adequate cooling water flow to the RCP thermal barriers into the plant operating procedures which establish the required flow.
On a loss of seal injection, the procedure did not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function.
The inspectors found that the procedurally established nominal flow band would have assured adequate cooling of the RCP thermal barriers for the highest CCW supply temperature recorded over the previous year.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because the operating procedure used to set the flow rate of cooling water to the RCP thermal barriers was not adequate to make certain that sufficient cooling water was available to assure the components could perform their design function.
(Section 1R15)
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH TESTING TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, NCV of 10 CFR 50 Appendix B, Criterion XI, Test Control, in that, Entergy did not establish appropriate testing to assure adequate component cooling water (CCW) flow to the reactor coolant pump thermal barriers. Specifically no preventive maintenance activities or functional checks were conducted for the individual flow meters. It was determined that the rotameters on 21 and 23 RCP were not indicating correctly and that actual CCW flow to the thermal barrier heat exchangers was less that the design requirements for CCW temperature.
Entergy entered this issue into their corrective action program (CR-IP2-2007-00783 and 00955), adjusted individual cooling water flow within the nominal band using ultrasonic flow meters, wrote work orders to replace the faulty flow meters, and is conducting an evaluation to determine the appropriate test requirements for the flow indicators.
 
This inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergys test program did not assure that all testing required to demonstrate that the RCP thermal barriers will perform satisfactorily in service because no testing was performed to ensure the accuracy of the individual flow meters used to establish the required cooling water flow. Consequently, it was identified that two individual flow indicators did not read correctly and the CCW flow to two RCPs was not sufficient to assure adequate cooling in the event that seal water was lost based on the flow requirements established in design calculations. On a loss of seal injection, the cooling water flow would not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function. (Section 1R15)
Inspection Report# : 2007002 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE RISK ASSESSMENT FOR 21 MBFP STEAM INLET VALVE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR),
Part 50.65(a)(4), because Entergy did not adequately assess and manage the risk of on-line maintenance activities while operating with a degraded steam inlet valve on one of Entergys two main boiler feed pumps (MBFP).
Specifically, from November 16 through 21, 2006, the degraded condition of the 21 MBFP increased the likelihood of a reactor trip, but was not assessed or included in the plants on-line risk model. Entergy entered this issue into their corrective action program and properly assessed 21 MBFP risk on November 21, 2006.
The inspectors determined that this finding was more than minor because Entergy failed to consider risk significant structures, systems, components, and support systems that were unavailable during the performance of on-line maintenance. Specifically, Entergy failed to assess the increase in online risk from the increased likelihood of a reactor trip due to the 21 MBFP degraded condition. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," and determined that this finding was of very low safety significance because the finding resulted in an increase in the incremental core damage probability of less than 1x10-6 (actual increase was approximately 2x10-8).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not provide complete and accurate procedures, in that, the online risk assessment procedure did not require degraded equipment that impacted risk to be assessed or managed.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE OPERATING PROCEDURES FOR LOSS OF BOTH HEATER DRAIN TANK PUMPS A Green self-revealing finding was identified because Entergy failed to develop adequate procedures for governing the response to a loss of both heater drain tank pumps and to an approaching rod insertion limit (RIL) alarm condition.
Specifically, the procedure governing operator actions during a loss of heater drain tank pumps did not specify for the operators to reset the steam dumps following the rapid downpower. The alarm response procedure for the approaching rod insertion limit condition directed the operators to place the rod control system in manual to stop further automatic inward rod motion. This impacted operators ability to add negative reactivity and control the transient. Entergy entered these procedural deficiencies into their corrective action program and is evaluating the appropriate steps to correct the procedural deficiencies.
 
The inspectors determined that this finding is greater than minor because it is associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the procedural inadequacies complicated operator actions to a rapid downpower, resulted in a manual reactor trip when the operators determined that they did not have sufficient control of the transient, and could impact other accident sequences requiring negative reactivity addition. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that plant operating procedures were adequate to ensure operators could appropriately respond to a rapid downpower transient.
Inspection Report# : 2006004 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE PROCEDURE FOR CALIBRATING THE STEAM DUMP LOSS OF LOAD CONTROLLER A Green self-revealing finding was identified because Entergy failed to develop an accurate procedure for calibration of the steam dump loss of load controller. This resulted in the steam dumps failing to operate properly during a plant transient, complicating operator response, and leading to a manual reactor trip. Following identification of the issue, Entergy entered the issue into the corrective action program, corrected the procedural deficiency, and re-calibrated the controller.
The inspectors determined that this finding is greater than minor because it is associated with the Procedural Quality attribute of the Initiating events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the inadequacy in Entergy's calibration procedure caused the steam dumps to operate improperly during a plant transient and contributed to a reactor trip. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be available. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for calibration of the steam dump loss of load controller was accurate, in that, it specified incorrect settings for the controller.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL ASSOCIATED WITH VORTEXING AND NET POSITIVE SUCTION HEAD CALCULATIONS The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not ensure adequate suction submergence for the three safety injection (SI) pumps by not properly translating vortex and net positive suction head (NPSH) design parameters into calculations relative to reactor water storage tank (RWST) level. Specifically, Entergy used a non-conservative method to calculate the level required to prevent pump vortexing, and used a non-conservative RWST level value for determining available NPSH for the SI pumps. Entergy entered the issue into their corrective action program and revised the affected calculations.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of the SI pumps, even though the pumps were ultimately shown to be operable. The finding is associated with the design
 
control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the significance determination process (SDP), documented in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," because it was a design deficiency that did not result in a loss of SI system operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DIFFERENTIAL PRESSURE VALUE USED FOR MOV 746 AND MOV 747 TONENSURE VALVE CAPABILITY The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not accurately incorporate design parameters into valve thrust calculations for motor operated valve (MOV) 746 and MOV 747. Specifically, Entergy used an incorrect and non-conservative differential pressure in the calculations for MOV 746 and MOV 747, which were developed to verify that the valves could develop sufficient thrust to open under postulated design basis conditions. Additionally, an incorrect equation was used in determining the reduction in motor torque due to degraded voltage conditions. Entergy entered the issue into their corrective action program and revised the affected calculations using the correct information.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of MOV 746 and MOV 747. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it was a design deficiency that did not to result in a loss of MOV 746 and MOV 747 operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR ENVIRONMENTAL EFFECTS TO ENSURE THE AVAILABILITY OF THE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP OPERATION The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not establish adequate design control measures to ensure the availability of the turbine driven auxiliary feedwater pump (TDAFWP) during a postulated loss-of-offsite power (LOOP) event. Under certain LOOP situations, the team determined that the TDAFWP steam supply could be inadvertently isolated because of inadequate calculations and procedures for limiting the AFWP room temperature rise. Specifically, a calculation to determine the auxiliary feedwater pump (AFWP) room temperature rise during a LOOP did not include heat input from the TDAFWP. Further, actions that could limit the rise in AFWP room temperature and prevent the inadvertent isolation of the TDAFW pump (opening an AFWP room roll-up door or promptly restoring forced ventilation) were not included in procedures. Entergy entered this issue into their corrective action program, implemented immediate compensatory actions, and revised AFWP room temperature rise calculations.
The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of safety function of the TDAFWP (single train) for greater than its 72 hour technical specification allowed outage time, based on the teams review and assessment of site ambient temperature data over the last year.
Inspection Report# : 2007007 (pdf)
 
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY MONITOR GAS TURBINE SYSTEM PERFORMANCE AS REQUIRED BY THE MAINTENANCE RULE The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50.65 (a)(1), the Maintenance Rule, in that, Entergy failed to monitor the gas turbine (GT) system in a manner that provided reasonable assurance that the system could perform its intended safety function. Specifically, Entergy did not establish appropriate GT reliability goals, and therefore did not take corrective actions, when GT-1 had exceeded these goals for maintenance preventable functions failures (MPFF). In addition, Entergy did not properly classify repeat MPFFs, which resulted in a similar failure to take corrective actions as required. This resulted in additional GT-1 out of service time that would not have happened if appropriate actions had been taken. Entergy entered this issue into their corrective action program and lowered the allowable goal for MPFFs, and revised the GT-1 (a)(1) action plan to improve reliability.
The finding is more than minor because appropriate GT reliability goals were not established commensurate with safety and appropriate corrective actions were not taken when goals were not met. This finding is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, which considered that the additional GT-1 out of service time due to this issue could be as much as three days. The finding has a cross-cutting aspect in the area of human performance because Entergy did not adequately ensure procedures were complete, accurate, and up-to-date. Specifically, procedure ENN-DC-171, Maintenance Rule Monitoring, did not provide steps to discriminate between the classification of an initial design deficiency and further failures due to the same condition, resulting in mis-classifying several GT functional failures.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT DEGRADED GAS TURBINE 1 RELIABILITY The team identified a finding of very low safety significance involving Entergy procedure, EN-LI-102, Corrective Action Process, in that, Entergy failed to take corrective actions to address degraded GT-1 reliability. This resulted in a two and one half day time period in January 2007 when GT-1 and GT-3 were simultaneously inoperable because, after GT-3 was made inoperable for planned maintenance activities, GT-1 was subsequently found to be inoperable.
Specifically, the reliability of GT-1 declined from an average of 75% for 2005 and the first 10 months of 2006, to 50% for the three months from November 2006 to January 2007; however, Entergy did not take actions to correct this degraded reliability. Entergy entered this issue into their corrective action program and developed an action plan to address GT reliability issues.
The issue is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, assuming that both GT-1 and GT-3 were unavailable for the two and one half days, due to this issue. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not correct degraded reliability of GT-1, resulting in having GT-1 and GT-3 simultaneously inoperable.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE STATION BATTERY CAPACITY TESTING FOR DEGRADATION MONITORING The team identified a finding of very low safety significance (Green) involving a non-cited violation of Technical Specification 3.8.6.6, in that, Entergy did not perform station battery capacity testing in accordance with IEEE
 
Standard 450-1995 (related to battery maintenance and testing). Specifically, Entergy procedurally terminated battery capacity testing at the rated discharge time (four hours), before reaching the minimum voltage, as specified by IEEE Standard 450-1995. This prevented accurate quantitative measurement of capacity degradation and identification of the need to conduct potential accelerated battery testing, as specified by both IEEE Standard 450-1995 and the technical specifications, if battery capacity drops by more than 10% relative to the previous test. Entergy entered the issue into their corrective action program and performed calculations using past test data, which demonstrated that the capacities of station batteries had not degraded more than 10%.
This issue is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of station battery safety function, based upon the teams verification of Entergys calculations.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION FOR HIGH INTER-TIER BATTERY RESISTANCES The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not take effective corrective actions for a condition adverse to quality concerning out-of-tolerance inter-tier resistances on the No. 21 station battery. Specifically, after repeated failures of the No. 21 station battery inter-tier resistance testing, vendor and IEEE Standard 450-1995 recommended corrective actions were not taken to correct the adverse out-of-tolerance resistance trend. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated that the voltage drop due to the as-found resistance of the inter-tier connections was small and did not impact No. 21 battery operability.
This issue is more than minor because if it was left uncorrected, it would have become a more significant safety concern. Specifically, high resistance connections in a battery that is loaded during accident conditions can cause localized heating and can cause permanent damage to the battery. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 21 station battery safety function, based upon the teams verification of Entergys revised calculations. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not take effective corrective actions to address the adverse trend of out-of-tolerance inter-tier resistances.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS FOR DECREASE IN BATTERY MARGIN The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not promptly identify and correct a condition adverse to quality, with respect to known errors in the No. 23 station battery design calculations. Specifically, Entergy did not recognize at the appropriate time the need to write a condition report, perform an operability determination, or place controls on the use of the No. 23 battery design calculations when errors were discovered in the No. 23 battery design calculations that significantly lowered the battery capacity margin. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated No. 23 station battery operability through the next refueling outage, based on the calculated margin and conservatisms available.
This issue is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 23 station battery safety function, based upon the teams verification of Entergys revised calculations.
 
The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy failed to promptly identify the decrease in margin found in the No. 23 battery design calculations of record.
Inspection Report# : 2007007 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO CORRECT A DEGRADED CONDITION WHICH IMPACTED GAS TURBINE #1 RELIABILITY AND AVAILABILITY The inspectors identified a Green finding, in that, Entergy's corrective actions were inadequate to resolve a deficiency associated with the gas turbine 1 (GT-1) starting diesel. This deficiency was identified following a failure of GT-1 to start on February 7, 2005, and resulted in three subsequent failures. A corrective action was written to correct the deficient condition following the initial failure and was closed on June 22, 2005, with no actions taken based on a senior management decision to cancel preventive maintenance activities on the gas turbines due to pending system retirement. Entergy entered this issue into their corrective action program and installed a modification to the coolant system to prevent further trips due to this condition.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it impacted GT-1 reliability, in that, the deficiency resulted in multiple failures to start on demand after the condition was identified and the action to correct the condition was closed without being implemented. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that a Phase 2 evaluation was required because the finding represented an actual loss of safety function of a non-Technical Specification required train of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours. The inspectors used the Risk-Informed Inspection Notebook for Indian Point Nuclear Generating Unit 2, to conduct the Phase 2 evaluation.
The inspectors determined that 65 hours of unavailability were caused by the additional failures of GT-1 due to the starting diesel coolant system deficiency. The inspectors conservatively equated this cumulative unavailability time to the total exposure time and used an initiating events likelihood of less than three days. The Phase 2 approximation yielded a result of very low safety significance (Green).
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available and adequate to assure reliable operation of GT-1.
Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the gas turbine system.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY A DEGRADED CONDITION OF AN AUXILIARY FEED WATER CHECK VALVE IN THE CORRECTIVE ACTION PROGRAM The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program. Specifically, upon inspection in September 2006, the gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation. Gasket damage was noted in work orders during internal valve inspections of BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in the corrective action program. Consequently, the problem was not evaluated and corrected prior to reassembly of the valve. Entergy entered this issue into the corrective action program, evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
 
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
Inspection Report# : 2006006 (pdf)
Significance:        Dec 05, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE EVALUATION OF LEAKING 22 STEAM GENERATOR LOW FLOW BYPASS VALVE FCV-427L A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage into the 22 steam generator. During the Indian Point Unit 2 reactor trip on August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively and resulted in an uncontrolled rise in 22 steam generator level; operator response to isolate feedwater to the steam generator in accordance with emergency operating procedures; and automatic actuation of the feedwater isolation system. The excessive leakage condition into the 22 steam generator was identified on April 4, 2006, prior to Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior to the reactor trip on August 23, 2006. This issue was entered into the corrective action program, and FCV-427L was repaired and retested satisfactorily.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspectors evaluated the significance of the finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolutions addressed the causes and extent of condition of the problem.
Inspection Report# : 2006006 (pdf)
Barrier Integrity Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MOVE CONTAINMENT HYDROGEN ANALYZERS TO 10 CFR 50.65 (A)(1) STATUS The inspectors identified a Green, NCV of 10 CFR 50.65(a)(2) because Entergy did not demonstrate that the performance or condition of the containment hydrogen monitoring system was being effectively controlled through the performance of appropriate preventive maintenance such that the system remained capable of performing its intended function. The inspectors identified that both channels of the containment hydrogen/oxygen (H2/O2) analyzers had been out of service since September 7, 2006, due to compressor seal leakage. The inspectors determined that the H2/O2 analyzers are within the scope of Entergys Maintenance Rule program since they are used in the emergency operating procedures. The inspectors noted that, based on the significant unavailability time of both trains, the system should have been in 10 CFR 50.65(a)(1) status with an action plan to improve system performance back to an (a)(2) status. Entergy entered this issue into their corrective action program and changed the priority of the work orders to perform repairs on the H2/O2 analyzers.
This inspectors determined that this finding affected the Barrier Integrity cornerstone and was more than minor since
 
it was similar to Example 7.b in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, Entergy failed to demonstrate effective control of the performance of the H2/O2 analyzers and did not place the system in (a)(1) status.
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The finding required further evaluation through IMC 0609, Appendix H, Containment Integrity Significance Determination Process, because it resulted in an actual reduction in the defense-in-depth for the hydrogen control function of the reactor containment. The inspectors determined that this finding was of very low safety significance because it did not affect core damage frequency and the H2/O2 analyzers are not important to large early release frequency.
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available to assure reliable operation of the H2/O2 analyzers.
Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the containment hydrogen monitoring system. (Section 4OA2)
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONTAINMENT CLOSURE EQUIPMENT The inspectors identified a Severity Level IV NCV of 10 CFR 50.59, Changes, Tests and Experiments, for failure to obtain a license amendment pursuant to 10 CFR 50.90 prior implementing a change to alter the requirements of a shutdown fission product barrier. The inspectors reviewed Safety Evaluation 04-0732-MD-00-RE R1, Installation of a Temporary Roll-up Door on the Containment Equipment Hatch, to determine if the conclusion that a licensee amendment was not required was correct. Entergy concluded that the roll-up door was equivalent to the closure plate and, therefore, adequate to close containment as required by the action statement. The inspectors found that the door was not designed to be air-tight; therefore, any radioactive release inside containment would bypass the roll-up door.
The inspectors concluded that the roll-up door did not meet the design or licensing basis of the closure plate as described in the Updated Final Safety Analysis Report (UFSAR) and previously approved license amendments.
Consequently, Entergy incorrectly concluded that a license amendment pursuant to 50.90 was not required prior to implementing the change. Entergy entered the issue into their corrective action program to evaluate and correct.
The inspectors determined that Entergy changed the requirements for the shutdown fission product barrier (containment) prior to receiving NRC approval. As a result, traditional enforcement was used to evaluate the issue because the deficiency affected the NRC's ability to perform its regulatory function. The severity level of the violation was determined to be Severity Level IV in accordance with example D.5 of Supplement 1 of the NRC Enforcement Policy. Additionally, the issue was determined to be of very low safety significance (Green) based on the low decay heat levels at the time the roll-up door was credited in accordance with the significance determination process described in Inspection Manual Chapter (IMC) 0609 Appendix H, Containment Integrity.
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Significance:        Mar 31, 2007 Identified By: NRC Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR FAILURE TO APPROPRIATELY MONITOR SERVICE WATER INTAKE BAY LEVEL The inspectors identified a Green finding because Entergy failed to take adequate corrective actions for an issue associated with monitoring of service water intake bay level. This deficiency could have prevented identification of entry conditions for an emergency action level. Entergy entered this issue into the corrective action program as CR IP3-2007-00453, and initiated several corrective actions, including plans for enhanced monitoring of service water bay levels, backwashing of trash racks, procedural upgrades, correction of service water bay level instrumentation modification installation, development of modifications for enhanced service water level monitoring equipment, and
 
enhanced inspection and cleaning of intake structure trash racks.
The inspectors determined that this finding was more than minor because it was associated with the Emergency Preparedness cornerstone attribute of facilities and equipment; and, it affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, inadequate monitoring of service water intake bay level could have resulted in failure to declare a notification of unusual event (UE). The inspectors reviewed the EAL entry criteria and determined that this performance deficiency did not affect Entergys ability to declare any event higher than a UE.
The inspectors evaluated this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, and determined that it was of very low safety significance because the declaration of a UE based on low service water bay level could have been missed or delayed, consistent with the example provided in the appendix.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement effective corrective actions for a previously identified issue associated with inadequate monitoring of service water intake bay level. (Section 1R17)
Inspection Report# : 2007002 (pdf)
Occupational Radiation Safety Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO SURVEY AND PROVIDE ACCESS TO AN UNPOSTED HIGH RADIATION AREA A Green, self-revealing NCV of 10 CFR 20.1501 with respect to 10 CFR 20.1902(b) was identified, in that, Entergy failed to survey radiological condition changes after a plant manipulation that was likely to cause a change in radiological conditions, and this led to the failure to post a plant area as a high radiation area. As a result, two workers were allowed access to an unsurveyed and unposted high radiation area.
The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of exposure control and affected the cornerstone objective, because not establishing radiological conditions and commensurate controls after changing plant radiological conditions prior to allowing access to the affected areas can cause increased personnel exposure. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that it was of very low safety significance (Green) because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. This issue was entered into Entergy's corrective action program and training was provided to the radiation protection staff.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not use a conservative assumption in the decision-making process, in that, the watch radiation protection technician did not question the radiological conditions of the pipe chase area after a change of plant conditions had occurred and did not require a survey of the pipe chase area before authorizing access to personnel.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding UNIT 2 CONTAINMENT SUMP STRAINER MODIFICATION COLLECTIVE EXPOSURE OVERRUNS DUE TO INADEQUATE MOD PREPARATION A self-revealing finding was identified that involved inadequate modification planning and construction preparations relative to a Unit 2 containment sump strainer modification that resulted in significant unplanned collective exposure (93.7 person-rem compared to a work activity estimate of 10.9 person-rem). Specifically, the actual job site conditions
 
for installation of the containment sump modification were not adequately evaluated with respect to the radiological impact of increased occupancy in high dose rate work areas. This unplanned additional in-field high radiation work resulted in significant unintended exposure that could have been avoided. This issue was entered into Entergy's corrective action program so that lessons learned could be incorporated into the Unit 3 containment sump modification.
The inspectors determined that this finding was more than minor because it was similar to examples 6.a and 6.b of IMC 0612, Appendix E, "Examples of Minor Issues," in that, the issue involved actual collective exposure greater than 5 person-rem and was greater than 50 percent above the estimated or intended exposure; and the majority of the dose overrun was due to activities within Entergy's control. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance (Green) because it involved an ALARA planning issue, and the 3-year rolling average collective dose for Unit 2 was less than 135 person-rem (73 person-rem average annual exposure for 2003 through 2005).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not adequately incorporate job site conditions in the work control planning process.
Inspection Report# : 2006005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance:        Dec 05, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO ENTER SAFETY CULTURE ASSESSMENT RESULTS INTO CORRECTIVE ACTION PROGRAM The NRC inspectors identified a finding when Entergy failed to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for the adverse conditions identified in the 2006 Safety Culture Assessment.
Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner. The contractor who performed the independent safety culture assessment presented the site specific results to Entergy management in June 2006. The negative responses and declining trends identified in the assessment constituted adverse conditions that should have been entered into the corrective action program. At the time of the inspection, Entergy had not initiated condition reports for the assessment results. Consequently, the results had not been fully evaluated to understand the causes and identify appropriate actions to address the identified issues.
Additionally, organizations identified by the contractor as needing management attention had not developed departmental action plans at the time of the inspection. Entergy entered this issue into the corrective action program and initiated a learning organization condition report to track development and implementation of action plans to address the assessment results.
The inspectors determined that the finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue,
 
increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance. The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify issues with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner.
Inspection Report# : 2006006 (pdf)
Last modified : August 24, 2007
 
Indian Point 2 3Q/2007 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INCORPORATE DESIGN BASIS INFORMATION INTO PROCEDURES TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not appropriately incorporate design requirements into an operating procedure used to establish adequate component cooling water (CCW) flow to the reactor coolant pump (RCP) thermal barriers.
Specifically, the flow specification in the CCW operating procedure did not incorporate the calculated design flow requirements to bound allowable CCW temperature limits. Entergy entered this issue into their corrective action program and will be evaluating the flow requirements specified in procedure 2-SOP-4.1.2, Component Cooling Water System Operation, to ensure that they bound the allowed plant operating limits.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergy did not incorporate design flow requirements necessary to assure adequate cooling water flow to the RCP thermal barriers into the plant operating procedures which establish the required flow.
On a loss of seal injection, the procedure did not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function.
The inspectors found that the procedurally established nominal flow band would have assured adequate cooling of the RCP thermal barriers for the highest CCW supply temperature recorded over the previous year.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because the operating procedure used to set the flow rate of cooling water to the RCP thermal barriers was not adequate to make certain that sufficient cooling water was available to assure the components could perform their design function.
(Section 1R15)
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH TESTING TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, NCV of 10 CFR 50 Appendix B, Criterion XI, Test Control, in that, Entergy did not establish appropriate testing to assure adequate component cooling water (CCW) flow to the reactor coolant pump thermal barriers. Specifically no preventive maintenance activities or functional checks were conducted for the individual flow meters. It was determined that the rotameters on 21 and 23 RCP were not indicating correctly and that actual CCW flow to the thermal barrier heat exchangers was less that the design requirements for CCW temperature.
Entergy entered this issue into their corrective action program (CR-IP2-2007-00783 and 00955), adjusted individual cooling water flow within the nominal band using ultrasonic flow meters, wrote work orders to replace the faulty flow meters, and is conducting an evaluation to determine the appropriate test requirements for the flow indicators.
 
This inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergys test program did not assure that all testing required to demonstrate that the RCP thermal barriers will perform satisfactorily in service because no testing was performed to ensure the accuracy of the individual flow meters used to establish the required cooling water flow. Consequently, it was identified that two individual flow indicators did not read correctly and the CCW flow to two RCPs was not sufficient to assure adequate cooling in the event that seal water was lost based on the flow requirements established in design calculations. On a loss of seal injection, the cooling water flow would not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function. (Section 1R15)
Inspection Report# : 2007002 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE RISK ASSESSMENT FOR 21 MBFP STEAM INLET VALVE The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR),
Part 50.65(a)(4), because Entergy did not adequately assess and manage the risk of on-line maintenance activities while operating with a degraded steam inlet valve on one of Entergys two main boiler feed pumps (MBFP).
Specifically, from November 16 through 21, 2006, the degraded condition of the 21 MBFP increased the likelihood of a reactor trip, but was not assessed or included in the plants on-line risk model. Entergy entered this issue into their corrective action program and properly assessed 21 MBFP risk on November 21, 2006.
The inspectors determined that this finding was more than minor because Entergy failed to consider risk significant structures, systems, components, and support systems that were unavailable during the performance of on-line maintenance. Specifically, Entergy failed to assess the increase in online risk from the increased likelihood of a reactor trip due to the 21 MBFP degraded condition. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," and determined that this finding was of very low safety significance because the finding resulted in an increase in the incremental core damage probability of less than 1x10-6 (actual increase was approximately 2x10-8).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not provide complete and accurate procedures, in that, the online risk assessment procedure did not require degraded equipment that impacted risk to be assessed or managed.
Inspection Report# : 2006005 (pdf)
Mitigating Systems Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL ASSOCIATED WITH VORTEXING AND NET POSITIVE SUCTION HEAD CALCULATIONS The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not ensure adequate suction submergence for the three safety injection (SI) pumps by not properly translating vortex and net positive suction head (NPSH) design parameters into
 
calculations relative to reactor water storage tank (RWST) level. Specifically, Entergy used a non-conservative method to calculate the level required to prevent pump vortexing, and used a non-conservative RWST level value for determining available NPSH for the SI pumps. Entergy entered the issue into their corrective action program and revised the affected calculations.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of the SI pumps, even though the pumps were ultimately shown to be operable. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the significance determination process (SDP), documented in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," because it was a design deficiency that did not result in a loss of SI system operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DIFFERENTIAL PRESSURE VALUE USED FOR MOV 746 AND MOV 747 TONENSURE VALVE CAPABILITY The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not accurately incorporate design parameters into valve thrust calculations for motor operated valve (MOV) 746 and MOV 747. Specifically, Entergy used an incorrect and non-conservative differential pressure in the calculations for MOV 746 and MOV 747, which were developed to verify that the valves could develop sufficient thrust to open under postulated design basis conditions. Additionally, an incorrect equation was used in determining the reduction in motor torque due to degraded voltage conditions. Entergy entered the issue into their corrective action program and revised the affected calculations using the correct information.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of MOV 746 and MOV 747. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it was a design deficiency that did not to result in a loss of MOV 746 and MOV 747 operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR ENVIRONMENTAL EFFECTS TO ENSURE THE AVAILABILITY OF THE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP OPERATION The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not establish adequate design control measures to ensure the availability of the turbine driven auxiliary feedwater pump (TDAFWP) during a postulated loss-of-offsite power (LOOP) event. Under certain LOOP situations, the team determined that the TDAFWP steam supply could be inadvertently isolated because of inadequate calculations and procedures for limiting the AFWP room temperature rise. Specifically, a calculation to determine the auxiliary feedwater pump (AFWP) room temperature rise during a LOOP did not include heat input from the TDAFWP. Further, actions that could limit the rise in AFWP room temperature and prevent the inadvertent isolation of the TDAFW pump (opening an AFWP room roll-up door or promptly restoring forced ventilation) were not included in procedures. Entergy entered this issue into their corrective action program, implemented immediate compensatory actions, and revised AFWP room temperature rise calculations.
The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems
 
that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of safety function of the TDAFWP (single train) for greater than its 72 hour technical specification allowed outage time, based on the teams review and assessment of site ambient temperature data over the last year.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY MONITOR GAS TURBINE SYSTEM PERFORMANCE AS REQUIRED BY THE MAINTENANCE RULE The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50.65 (a)(1), the Maintenance Rule, in that, Entergy failed to monitor the gas turbine (GT) system in a manner that provided reasonable assurance that the system could perform its intended safety function. Specifically, Entergy did not establish appropriate GT reliability goals, and therefore did not take corrective actions, when GT-1 had exceeded these goals for maintenance preventable functions failures (MPFF). In addition, Entergy did not properly classify repeat MPFFs, which resulted in a similar failure to take corrective actions as required. This resulted in additional GT-1 out of service time that would not have happened if appropriate actions had been taken. Entergy entered this issue into their corrective action program and lowered the allowable goal for MPFFs, and revised the GT-1 (a)(1) action plan to improve reliability.
The finding is more than minor because appropriate GT reliability goals were not established commensurate with safety and appropriate corrective actions were not taken when goals were not met. This finding is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, which considered that the additional GT-1 out of service time due to this issue could be as much as three days. The finding has a cross-cutting aspect in the area of human performance because Entergy did not adequately ensure procedures were complete, accurate, and up-to-date. Specifically, procedure ENN-DC-171, Maintenance Rule Monitoring, did not provide steps to discriminate between the classification of an initial design deficiency and further failures due to the same condition, resulting in mis-classifying several GT functional failures.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT DEGRADED GAS TURBINE 1 RELIABILITY The team identified a finding of very low safety significance involving Entergy procedure, EN-LI-102, Corrective Action Process, in that, Entergy failed to take corrective actions to address degraded GT-1 reliability. This resulted in a two and one half day time period in January 2007 when GT-1 and GT-3 were simultaneously inoperable because, after GT-3 was made inoperable for planned maintenance activities, GT-1 was subsequently found to be inoperable.
Specifically, the reliability of GT-1 declined from an average of 75% for 2005 and the first 10 months of 2006, to 50% for the three months from November 2006 to January 2007; however, Entergy did not take actions to correct this degraded reliability. Entergy entered this issue into their corrective action program and developed an action plan to address GT reliability issues.
The issue is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, assuming that both GT-1 and GT-3 were unavailable for the two and one half days, due to this issue. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not correct degraded reliability of GT-1, resulting in having GT-1 and GT-3 simultaneously inoperable.
Inspection Report# : 2007007 (pdf)
 
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE STATION BATTERY CAPACITY TESTING FOR DEGRADATION MONITORING The team identified a finding of very low safety significance (Green) involving a non-cited violation of Technical Specification 3.8.6.6, in that, Entergy did not perform station battery capacity testing in accordance with IEEE Standard 450-1995 (related to battery maintenance and testing). Specifically, Entergy procedurally terminated battery capacity testing at the rated discharge time (four hours), before reaching the minimum voltage, as specified by IEEE Standard 450-1995. This prevented accurate quantitative measurement of capacity degradation and identification of the need to conduct potential accelerated battery testing, as specified by both IEEE Standard 450-1995 and the technical specifications, if battery capacity drops by more than 10% relative to the previous test. Entergy entered the issue into their corrective action program and performed calculations using past test data, which demonstrated that the capacities of station batteries had not degraded more than 10%.
This issue is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of station battery safety function, based upon the teams verification of Entergys calculations.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION FOR HIGH INTER-TIER BATTERY RESISTANCES The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not take effective corrective actions for a condition adverse to quality concerning out-of-tolerance inter-tier resistances on the No. 21 station battery. Specifically, after repeated failures of the No. 21 station battery inter-tier resistance testing, vendor and IEEE Standard 450-1995 recommended corrective actions were not taken to correct the adverse out-of-tolerance resistance trend. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated that the voltage drop due to the as-found resistance of the inter-tier connections was small and did not impact No. 21 battery operability.
This issue is more than minor because if it was left uncorrected, it would have become a more significant safety concern. Specifically, high resistance connections in a battery that is loaded during accident conditions can cause localized heating and can cause permanent damage to the battery. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 21 station battery safety function, based upon the teams verification of Entergys revised calculations. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not take effective corrective actions to address the adverse trend of out-of-tolerance inter-tier resistances.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS FOR DECREASE IN BATTERY MARGIN The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not promptly identify and correct a condition adverse to quality, with respect to known errors in the No. 23 station battery design calculations. Specifically, Entergy did not recognize at the appropriate time the need to write a condition report, perform an operability determination, or place controls on the use of the No. 23 battery design calculations when errors were discovered in the No. 23 battery design calculations that significantly lowered the battery capacity margin. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated No. 23 station battery operability through the next refueling outage, based on the calculated margin and conservatisms available.
 
This issue is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 23 station battery safety function, based upon the teams verification of Entergys revised calculations.
The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy failed to promptly identify the decrease in margin found in the No. 23 battery design calculations of record.
Inspection Report# : 2007007 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO CORRECT A DEGRADED CONDITION WHICH IMPACTED GAS TURBINE #1 RELIABILITY AND AVAILABILITY The inspectors identified a Green finding, in that, Entergy's corrective actions were inadequate to resolve a deficiency associated with the gas turbine 1 (GT-1) starting diesel. This deficiency was identified following a failure of GT-1 to start on February 7, 2005, and resulted in three subsequent failures. A corrective action was written to correct the deficient condition following the initial failure and was closed on June 22, 2005, with no actions taken based on a senior management decision to cancel preventive maintenance activities on the gas turbines due to pending system retirement. Entergy entered this issue into their corrective action program and installed a modification to the coolant system to prevent further trips due to this condition.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it impacted GT-1 reliability, in that, the deficiency resulted in multiple failures to start on demand after the condition was identified and the action to correct the condition was closed without being implemented. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined that a Phase 2 evaluation was required because the finding represented an actual loss of safety function of a non-Technical Specification required train of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours. The inspectors used the Risk-Informed Inspection Notebook for Indian Point Nuclear Generating Unit 2, to conduct the Phase 2 evaluation.
The inspectors determined that 65 hours of unavailability were caused by the additional failures of GT-1 due to the starting diesel coolant system deficiency. The inspectors conservatively equated this cumulative unavailability time to the total exposure time and used an initiating events likelihood of less than three days. The Phase 2 approximation yielded a result of very low safety significance (Green).
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available and adequate to assure reliable operation of GT-1.
Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the gas turbine system.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY A DEGRADED CONDITION OF AN AUXILIARY FEED WATER CHECK VALVE IN THE CORRECTIVE ACTION PROGRAM The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to quality associated with improper internal clearances on BFD-68, an auxiliary feedwater check valve, in the corrective action program. Specifically, upon inspection in September 2006, the gasket between the valve's body to bonnet seal was found over-crushed causing the gasket to partially unwind, potentially impacting valve operation. Gasket damage was noted in work orders during internal valve inspections of BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in the
 
corrective action program. Consequently, the problem was not evaluated and corrected prior to reassembly of the valve. Entergy entered this issue into the corrective action program, evaluated the condition, and conducted repairs to the valve to ensure the proper gasket crush was obtained.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
Inspection Report# : 2006006 (pdf)
Significance:        Dec 05, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation INADEQUATE EVALUATION OF LEAKING 22 STEAM GENERATOR LOW FLOW BYPASS VALVE FCV-427L A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage into the 22 steam generator. During the Indian Point Unit 2 reactor trip on August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively and resulted in an uncontrolled rise in 22 steam generator level; operator response to isolate feedwater to the steam generator in accordance with emergency operating procedures; and automatic actuation of the feedwater isolation system. The excessive leakage condition into the 22 steam generator was identified on April 4, 2006, prior to Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior to the reactor trip on August 23, 2006. This issue was entered into the corrective action program, and FCV-427L was repaired and retested satisfactorily.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspectors evaluated the significance of the finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that the finding was of very low safety significance because it was not a design or qualification deficiency; it did not result in the loss of a system safety function or a train safety function for greater than the Technical Specification Allowed Outage Time; and it did not screen as potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam generator such that the resolutions addressed the causes and extent of condition of the problem.
Inspection Report# : 2006006 (pdf)
Barrier Integrity Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MOVE CONTAINMENT HYDROGEN ANALYZERS TO 10 CFR 50.65 (A)(1) STATUS The inspectors identified a Green, NCV of 10 CFR 50.65(a)(2) because Entergy did not demonstrate that the performance or condition of the containment hydrogen monitoring system was being effectively controlled through the performance of appropriate preventive maintenance such that the system remained capable of performing its intended function. The inspectors identified that both channels of the containment hydrogen/oxygen (H2/O2) analyzers had been out of service since September 7, 2006, due to compressor seal leakage. The inspectors determined
 
that the H2/O2 analyzers are within the scope of Entergys Maintenance Rule program since they are used in the emergency operating procedures. The inspectors noted that, based on the significant unavailability time of both trains, the system should have been in 10 CFR 50.65(a)(1) status with an action plan to improve system performance back to an (a)(2) status. Entergy entered this issue into their corrective action program and changed the priority of the work orders to perform repairs on the H2/O2 analyzers.
This inspectors determined that this finding affected the Barrier Integrity cornerstone and was more than minor since it was similar to Example 7.b in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, Entergy failed to demonstrate effective control of the performance of the H2/O2 analyzers and did not place the system in (a)(1) status.
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The finding required further evaluation through IMC 0609, Appendix H, Containment Integrity Significance Determination Process, because it resulted in an actual reduction in the defense-in-depth for the hydrogen control function of the reactor containment. The inspectors determined that this finding was of very low safety significance because it did not affect core damage frequency and the H2/O2 analyzers are not important to large early release frequency.
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available to assure reliable operation of the H2/O2 analyzers.
Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the containment hydrogen monitoring system. (Section 4OA2)
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CONTAINMENT CLOSURE EQUIPMENT The inspectors identified a Severity Level IV NCV of 10 CFR 50.59, Changes, Tests and Experiments, for failure to obtain a license amendment pursuant to 10 CFR 50.90 prior implementing a change to alter the requirements of a shutdown fission product barrier. The inspectors reviewed Safety Evaluation 04-0732-MD-00-RE R1, Installation of a Temporary Roll-up Door on the Containment Equipment Hatch, to determine if the conclusion that a licensee amendment was not required was correct. Entergy concluded that the roll-up door was equivalent to the closure plate and, therefore, adequate to close containment as required by the action statement. The inspectors found that the door was not designed to be air-tight; therefore, any radioactive release inside containment would bypass the roll-up door.
The inspectors concluded that the roll-up door did not meet the design or licensing basis of the closure plate as described in the Updated Final Safety Analysis Report (UFSAR) and previously approved license amendments.
Consequently, Entergy incorrectly concluded that a license amendment pursuant to 50.90 was not required prior to implementing the change. Entergy entered the issue into their corrective action program to evaluate and correct.
The inspectors determined that Entergy changed the requirements for the shutdown fission product barrier (containment) prior to receiving NRC approval. As a result, traditional enforcement was used to evaluate the issue because the deficiency affected the NRC's ability to perform its regulatory function. The severity level of the violation was determined to be Severity Level IV in accordance with example D.5 of Supplement 1 of the NRC Enforcement Policy. Additionally, the issue was determined to be of very low safety significance (Green) based on the low decay heat levels at the time the roll-up door was credited in accordance with the significance determination process described in Inspection Manual Chapter (IMC) 0609 Appendix H, Containment Integrity.
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Significance:        Mar 31, 2007 Identified By: NRC Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR FAILURE TO APPROPRIATELY MONITOR SERVICE
 
WATER INTAKE BAY LEVEL The inspectors identified a Green finding because Entergy failed to take adequate corrective actions for an issue associated with monitoring of service water intake bay level. This deficiency could have prevented identification of entry conditions for an emergency action level. Entergy entered this issue into the corrective action program as CR IP3-2007-00453, and initiated several corrective actions, including plans for enhanced monitoring of service water bay levels, backwashing of trash racks, procedural upgrades, correction of service water bay level instrumentation modification installation, development of modifications for enhanced service water level monitoring equipment, and enhanced inspection and cleaning of intake structure trash racks.
The inspectors determined that this finding was more than minor because it was associated with the Emergency Preparedness cornerstone attribute of facilities and equipment; and, it affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, inadequate monitoring of service water intake bay level could have resulted in failure to declare a notification of unusual event (UE). The inspectors reviewed the EAL entry criteria and determined that this performance deficiency did not affect Entergys ability to declare any event higher than a UE.
The inspectors evaluated this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, and determined that it was of very low safety significance because the declaration of a UE based on low service water bay level could have been missed or delayed, consistent with the example provided in the appendix.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement effective corrective actions for a previously identified issue associated with inadequate monitoring of service water intake bay level. (Section 1R17)
Inspection Report# : 2007002 (pdf)
Occupational Radiation Safety Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO SURVEY AND PROVIDE ACCESS TO AN UNPOSTED HIGH RADIATION AREA A Green, self-revealing NCV of 10 CFR 20.1501 with respect to 10 CFR 20.1902(b) was identified, in that, Entergy failed to survey radiological condition changes after a plant manipulation that was likely to cause a change in radiological conditions, and this led to the failure to post a plant area as a high radiation area. As a result, two workers were allowed access to an unsurveyed and unposted high radiation area.
The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of exposure control and affected the cornerstone objective, because not establishing radiological conditions and commensurate controls after changing plant radiological conditions prior to allowing access to the affected areas can cause increased personnel exposure. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that it was of very low safety significance (Green) because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. This issue was entered into Entergy's corrective action program and training was provided to the radiation protection staff.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not use a conservative assumption in the decision-making process, in that, the watch radiation protection technician did not question the radiological conditions of the pipe chase area after a change of plant conditions had occurred and did not require a survey of the pipe chase area before authorizing access to personnel.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006
 
Identified By: Self-Revealing Item Type: FIN Finding UNIT 2 CONTAINMENT SUMP STRAINER MODIFICATION COLLECTIVE EXPOSURE OVERRUNS DUE TO INADEQUATE MOD PREPARATION A self-revealing finding was identified that involved inadequate modification planning and construction preparations relative to a Unit 2 containment sump strainer modification that resulted in significant unplanned collective exposure (93.7 person-rem compared to a work activity estimate of 10.9 person-rem). Specifically, the actual job site conditions for installation of the containment sump modification were not adequately evaluated with respect to the radiological impact of increased occupancy in high dose rate work areas. This unplanned additional in-field high radiation work resulted in significant unintended exposure that could have been avoided. This issue was entered into Entergy's corrective action program so that lessons learned could be incorporated into the Unit 3 containment sump modification.
The inspectors determined that this finding was more than minor because it was similar to examples 6.a and 6.b of IMC 0612, Appendix E, "Examples of Minor Issues," in that, the issue involved actual collective exposure greater than 5 person-rem and was greater than 50 percent above the estimated or intended exposure; and the majority of the dose overrun was due to activities within Entergy's control. The inspectors evaluated this finding using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined that the finding was of very low safety significance (Green) because it involved an ALARA planning issue, and the 3-year rolling average collective dose for Unit 2 was less than 135 person-rem (73 person-rem average annual exposure for 2003 through 2005).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not adequately incorporate job site conditions in the work control planning process.
Inspection Report# : 2006005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance:        Dec 05, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO ENTER SAFETY CULTURE ASSESSMENT RESULTS INTO CORRECTIVE ACTION PROGRAM The NRC inspectors identified a finding when Entergy failed to initiate condition reports in accordance with EN-LI-102, Corrective Action Process, for the adverse conditions identified in the 2006 Safety Culture Assessment.
Consequently, the adverse conditions were not evaluated and appropriate corrective actions were not identified in a timely manner. The contractor who performed the independent safety culture assessment presented the site specific results to Entergy management in June 2006. The negative responses and declining trends identified in the assessment constituted adverse conditions that should have been entered into the corrective action program. At the time of the inspection, Entergy had not initiated condition reports for the assessment results. Consequently, the results had not been fully evaluated to understand the causes and identify appropriate actions to address the identified issues.
 
Additionally, organizations identified by the contractor as needing management attention had not developed departmental action plans at the time of the inspection. Entergy entered this issue into the corrective action program and initiated a learning organization condition report to track development and implementation of action plans to address the assessment results.
The inspectors determined that the finding was more than minor because if left uncorrected it would become a more significant safety concern. Without appropriate action, the weaknesses in the safety culture onsite would continue, increasing the potential that safety issues would not receive the attention warranted by their significance. The finding was not suitable for SDP evaluation, but has been reviewed by NRC management and has been determined to be a finding of very low safety significance. The finding was not greater than very low safety significance because the inspectors did not identify any issues that were not raised which had an actual impact on plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not identify issues with the potential to impact nuclear safety in the corrective action process for evaluation and resolution in a timely manner.
Inspection Report# : 2006006 (pdf)
Last modified : December 07, 2007
 
Indian Point 2 4Q/2007 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INCORPORATE DESIGN BASIS INFORMATION INTO PROCEDURES TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not appropriately incorporate design requirements into an operating procedure used to establish adequate component cooling water (CCW) flow to the reactor coolant pump (RCP) thermal barriers.
Specifically, the flow specification in the CCW operating procedure did not incorporate the calculated design flow requirements to bound allowable CCW temperature limits. Entergy entered this issue into their corrective action program and will be evaluating the flow requirements specified in procedure 2-SOP-4.1.2, Component Cooling Water System Operation, to ensure that they bound the allowed plant operating limits.
The inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergy did not incorporate design flow requirements necessary to assure adequate cooling water flow to the RCP thermal barriers into the plant operating procedures which establish the required flow.
On a loss of seal injection, the procedure did not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function.
The inspectors found that the procedurally established nominal flow band would have assured adequate cooling of the RCP thermal barriers for the highest CCW supply temperature recorded over the previous year.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because the operating procedure used to set the flow rate of cooling water to the RCP thermal barriers was not adequate to make certain that sufficient cooling water was available to assure the components could perform their design function.
(Section 1R15)
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH TESTING TO ASSURE ADEQUATE COOLING WATER FLOW TO THE RCP THERMAL BARRIERS The inspectors identified a Green, NCV of 10 CFR 50 Appendix B, Criterion XI, Test Control, in that, Entergy did not establish appropriate testing to assure adequate component cooling water (CCW) flow to the reactor coolant pump thermal barriers. Specifically no preventive maintenance activities or functional checks were conducted for the individual flow meters. It was determined that the rotameters on 21 and 23 RCP were not indicating correctly and that actual CCW flow to the thermal barrier heat exchangers was less that the design requirements for CCW temperature.
Entergy entered this issue into their corrective action program (CR-IP2-2007-00783 and 00955), adjusted individual cooling water flow within the nominal band using ultrasonic flow meters, wrote work orders to replace the faulty flow meters, and is conducting an evaluation to determine the appropriate test requirements for the flow indicators.
 
This inspectors determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone; and, it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Entergys test program did not assure that all testing required to demonstrate that the RCP thermal barriers will perform satisfactorily in service because no testing was performed to ensure the accuracy of the individual flow meters used to establish the required cooling water flow. Consequently, it was identified that two individual flow indicators did not read correctly and the CCW flow to two RCPs was not sufficient to assure adequate cooling in the event that seal water was lost based on the flow requirements established in design calculations. On a loss of seal injection, the cooling water flow would not ensure that the heat removal capability was adequate to prevent a rise in seal temperature which would require the RCP to be stopped with a subsequent reactor trip. The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding was determined to be of very low safety significance because it would not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and would not have likely affected other mitigating systems resulting in a loss of their safety function. (Section 1R15)
Inspection Report# : 2007002 (pdf)
Mitigating Systems Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation Degraded 12 Fire Main Booster Pump Cell Fire Door The inspectors identified a non-cited violation (NCV) of License Condition 2.K., fire protection program, because Entergy failed to identify a degraded three-hour rated fire door on the east entrance of the 12 fire main booster pump room. The door was determined to be inoperable due to a misalignment, which prevented the door from fully closing.
Entergy entered this issue into their corrective action program, took immediate compensatory action, realigned the door, and ensured that it would fully close.
Inspection Report# : 2007004 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation PROCEDURE INADEQUATE TO ENSURE OPERABILITY OF SI PUMPS DURING VENTING The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not ensure that procedures associated with operation of the safety injection (SI) system during venting were appropriate to the circumstances. Specifically, procedure 2-PT-M108, RHR/SI [residual heat removal/safety injection] System Venting, did not have appropriate controls to ensure the safety injection piping and pumps remained operable during accident conditions. Entergy entered the issue into their corrective action program and revised the venting procedure to ensure operator actions are appropriately evaluated and credited to maintain operability of the system.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone; and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined this finding resulted in a loss of function of a single train of SI for approximately five minutes. Because the total inoperability time was less than the allowed outage time of 72 hours, and the finding is not potentially risk significant due to a seismic, flooding, or severe weather initiating event, this finding screens as very low safety significance (Green).
 
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that complete, accurate and up-to-date procedures were available. (H.2(c))
Inspection Report# : 2007004 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL ASSOCIATED WITH VORTEXING AND NET POSITIVE SUCTION HEAD CALCULATIONS The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not ensure adequate suction submergence for the three safety injection (SI) pumps by not properly translating vortex and net positive suction head (NPSH) design parameters into calculations relative to reactor water storage tank (RWST) level. Specifically, Entergy used a non-conservative method to calculate the level required to prevent pump vortexing, and used a non-conservative RWST level value for determining available NPSH for the SI pumps. Entergy entered the issue into their corrective action program and revised the affected calculations.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of the SI pumps, even though the pumps were ultimately shown to be operable. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the significance determination process (SDP), documented in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," because it was a design deficiency that did not result in a loss of SI system operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DIFFERENTIAL PRESSURE VALUE USED FOR MOV 746 AND MOV 747 TONENSURE VALVE CAPABILITY The team identified a finding of very low significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not accurately incorporate design parameters into valve thrust calculations for motor operated valve (MOV) 746 and MOV 747. Specifically, Entergy used an incorrect and non-conservative differential pressure in the calculations for MOV 746 and MOV 747, which were developed to verify that the valves could develop sufficient thrust to open under postulated design basis conditions. Additionally, an incorrect equation was used in determining the reduction in motor torque due to degraded voltage conditions. Entergy entered the issue into their corrective action program and revised the affected calculations using the correct information.
The finding is more than minor because the calculation deficiencies represented reasonable doubt on the operability of MOV 746 and MOV 747. The finding is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it was a design deficiency that did not to result in a loss of MOV 746 and MOV 747 operability, based upon the teams verification of Entergys revised calculations.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL FOR ENVIRONMENTAL EFFECTS TO ENSURE THE AVAILABILITY OF THE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP OPERATION
 
The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not establish adequate design control measures to ensure the availability of the turbine driven auxiliary feedwater pump (TDAFWP) during a postulated loss-of-offsite power (LOOP) event. Under certain LOOP situations, the team determined that the TDAFWP steam supply could be inadvertently isolated because of inadequate calculations and procedures for limiting the AFWP room temperature rise. Specifically, a calculation to determine the auxiliary feedwater pump (AFWP) room temperature rise during a LOOP did not include heat input from the TDAFWP. Further, actions that could limit the rise in AFWP room temperature and prevent the inadvertent isolation of the TDAFW pump (opening an AFWP room roll-up door or promptly restoring forced ventilation) were not included in procedures. Entergy entered this issue into their corrective action program, implemented immediate compensatory actions, and revised AFWP room temperature rise calculations.
The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of safety function of the TDAFWP (single train) for greater than its 72 hour technical specification allowed outage time, based on the teams review and assessment of site ambient temperature data over the last year.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY MONITOR GAS TURBINE SYSTEM PERFORMANCE AS REQUIRED BY THE MAINTENANCE RULE The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50.65 (a)(1), the Maintenance Rule, in that, Entergy failed to monitor the gas turbine (GT) system in a manner that provided reasonable assurance that the system could perform its intended safety function. Specifically, Entergy did not establish appropriate GT reliability goals, and therefore did not take corrective actions, when GT-1 had exceeded these goals for maintenance preventable functions failures (MPFF). In addition, Entergy did not properly classify repeat MPFFs, which resulted in a similar failure to take corrective actions as required. This resulted in additional GT-1 out of service time that would not have happened if appropriate actions had been taken. Entergy entered this issue into their corrective action program and lowered the allowable goal for MPFFs, and revised the GT-1 (a)(1) action plan to improve reliability.
The finding is more than minor because appropriate GT reliability goals were not established commensurate with safety and appropriate corrective actions were not taken when goals were not met. This finding is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, which considered that the additional GT-1 out of service time due to this issue could be as much as three days. The finding has a cross-cutting aspect in the area of human performance because Entergy did not adequately ensure procedures were complete, accurate, and up-to-date. Specifically, procedure ENN-DC-171, Maintenance Rule Monitoring, did not provide steps to discriminate between the classification of an initial design deficiency and further failures due to the same condition, resulting in mis-classifying several GT functional failures.
Inspection Report# : 2007007 (pdf)
Significance:        Feb 16, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT DEGRADED GAS TURBINE 1 RELIABILITY The team identified a finding of very low safety significance involving Entergy procedure, EN-LI-102, Corrective Action Process, in that, Entergy failed to take corrective actions to address degraded GT-1 reliability. This resulted in a two and one half day time period in January 2007 when GT-1 and GT-3 were simultaneously inoperable because, after GT-3 was made inoperable for planned maintenance activities, GT-1 was subsequently found to be inoperable.
Specifically, the reliability of GT-1 declined from an average of 75% for 2005 and the first 10 months of 2006, to
 
50% for the three months from November 2006 to January 2007; however, Entergy did not take actions to correct this degraded reliability. Entergy entered this issue into their corrective action program and developed an action plan to address GT reliability issues.
The issue is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 and Phase 2 of the SDP, assuming that both GT-1 and GT-3 were unavailable for the two and one half days, due to this issue. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not correct degraded reliability of GT-1, resulting in having GT-1 and GT-3 simultaneously inoperable.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE STATION BATTERY CAPACITY TESTING FOR DEGRADATION MONITORING The team identified a finding of very low safety significance (Green) involving a non-cited violation of Technical Specification 3.8.6.6, in that, Entergy did not perform station battery capacity testing in accordance with IEEE Standard 450-1995 (related to battery maintenance and testing). Specifically, Entergy procedurally terminated battery capacity testing at the rated discharge time (four hours), before reaching the minimum voltage, as specified by IEEE Standard 450-1995. This prevented accurate quantitative measurement of capacity degradation and identification of the need to conduct potential accelerated battery testing, as specified by both IEEE Standard 450-1995 and the technical specifications, if battery capacity drops by more than 10% relative to the previous test. Entergy entered the issue into their corrective action program and performed calculations using past test data, which demonstrated that the capacities of station batteries had not degraded more than 10%.
This issue is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of station battery safety function, based upon the teams verification of Entergys calculations.
Inspection Report# : 2007007 (pdf)
Significance:      Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION FOR HIGH INTER-TIER BATTERY RESISTANCES The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not take effective corrective actions for a condition adverse to quality concerning out-of-tolerance inter-tier resistances on the No. 21 station battery. Specifically, after repeated failures of the No. 21 station battery inter-tier resistance testing, vendor and IEEE Standard 450-1995 recommended corrective actions were not taken to correct the adverse out-of-tolerance resistance trend. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated that the voltage drop due to the as-found resistance of the inter-tier connections was small and did not impact No. 21 battery operability.
This issue is more than minor because if it was left uncorrected, it would have become a more significant safety concern. Specifically, high resistance connections in a battery that is loaded during accident conditions can cause localized heating and can cause permanent damage to the battery. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 21 station battery safety function, based upon the teams verification of Entergys revised calculations. The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not take effective corrective actions to address the adverse trend of out-of-tolerance inter-tier resistances.
Inspection Report# : 2007007 (pdf)
 
Significance:        Feb 16, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS FOR DECREASE IN BATTERY MARGIN The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not promptly identify and correct a condition adverse to quality, with respect to known errors in the No. 23 station battery design calculations. Specifically, Entergy did not recognize at the appropriate time the need to write a condition report, perform an operability determination, or place controls on the use of the No. 23 battery design calculations when errors were discovered in the No. 23 battery design calculations that significantly lowered the battery capacity margin. Entergy entered the issue into their corrective action program and performed calculations, which demonstrated No. 23 station battery operability through the next refueling outage, based on the calculated margin and conservatisms available.
This issue is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has very low safety significance, based on Phase 1 of the SDP, because it did not represent the loss of No. 23 station battery safety function, based upon the teams verification of Entergys revised calculations.
The finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy failed to promptly identify the decrease in margin found in the No. 23 battery design calculations of record.
Inspection Report# : 2007007 (pdf)
Barrier Integrity Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS TO REPAIR A DEGRADED SERVICE WATER FLOW INSTRUMENT The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, in that, Entergy did not implement timely corrective actions for a degraded condition associated with the 25 Containment Fan Cooler Unit (FCU) flow indicator. Specifically, the failure to take timely corrective actions for the degraded service water flow indicator for the 25 FCU, initially identified in October 2006, resulted in the inability to ensure that sufficient service water flow was available for the component to perform its intended function. Subsequently, it was identified that a reduced service water flow condition did exist. Entergy entered the issue into their corrective action program and implemented corrective actions to restore adequate indication of service water flow to the 25 FCU.
Entergy is evaluating maintenance practices to determine the appropriateness of a periodic blow-down of the transmitter impulse lines to prevent sediment buildup.
The inspectors determined that this finding was more than minor because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that the physical design barrier (containment) protects the public from radionuclide releases caused by accidents or events. This finding was evaluated using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance because, while it could impact late containment failure, it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the condition when initially identified. (P.1(c))
Inspection Report# : 2007004 (pdf)
 
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MOVE CONTAINMENT HYDROGEN ANALYZERS TO 10 CFR 50.65 (A)(1) STATUS The inspectors identified a Green, NCV of 10 CFR 50.65(a)(2) because Entergy did not demonstrate that the performance or condition of the containment hydrogen monitoring system was being effectively controlled through the performance of appropriate preventive maintenance such that the system remained capable of performing its intended function. The inspectors identified that both channels of the containment hydrogen/oxygen (H2/O2) analyzers had been out of service since September 7, 2006, due to compressor seal leakage. The inspectors determined that the H2/O2 analyzers are within the scope of Entergys Maintenance Rule program since they are used in the emergency operating procedures. The inspectors noted that, based on the significant unavailability time of both trains, the system should have been in 10 CFR 50.65(a)(1) status with an action plan to improve system performance back to an (a)(2) status. Entergy entered this issue into their corrective action program and changed the priority of the work orders to perform repairs on the H2/O2 analyzers.
This inspectors determined that this finding affected the Barrier Integrity cornerstone and was more than minor since it was similar to Example 7.b in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, Entergy failed to demonstrate effective control of the performance of the H2/O2 analyzers and did not place the system in (a)(1) status.
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The finding required further evaluation through IMC 0609, Appendix H, Containment Integrity Significance Determination Process, because it resulted in an actual reduction in the defense-in-depth for the hydrogen control function of the reactor containment. The inspectors determined that this finding was of very low safety significance because it did not affect core damage frequency and the H2/O2 analyzers are not important to large early release frequency.
The inspectors determined this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that equipment and resources were available to assure reliable operation of the H2/O2 analyzers.
Specifically, Entergy did not minimize long-standing equipment issues and maintenance deferrals associated with the containment hydrogen monitoring system. (Section 4OA2)
Inspection Report# : 2007002 (pdf)
Emergency Preparedness Significance:        Mar 31, 2007 Identified By: NRC Item Type: FIN Finding INADEQUATE CORRECTIVE ACTIONS FOR FAILURE TO APPROPRIATELY MONITOR SERVICE WATER INTAKE BAY LEVEL The inspectors identified a Green finding because Entergy failed to take adequate corrective actions for an issue associated with monitoring of service water intake bay level. This deficiency could have prevented identification of entry conditions for an emergency action level. Entergy entered this issue into the corrective action program as CR IP3-2007-00453, and initiated several corrective actions, including plans for enhanced monitoring of service water bay levels, backwashing of trash racks, procedural upgrades, correction of service water bay level instrumentation modification installation, development of modifications for enhanced service water level monitoring equipment, and enhanced inspection and cleaning of intake structure trash racks.
The inspectors determined that this finding was more than minor because it was associated with the Emergency Preparedness cornerstone attribute of facilities and equipment; and, it affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, inadequate monitoring of service water intake bay level could have resulted in failure to declare a notification of unusual event (UE). The inspectors reviewed the EAL entry criteria and determined that this performance deficiency did not affect Entergys ability to declare any event higher than a UE.
 
The inspectors evaluated this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, and determined that it was of very low safety significance because the declaration of a UE based on low service water bay level could have been missed or delayed, consistent with the example provided in the appendix.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not implement effective corrective actions for a previously identified issue associated with inadequate monitoring of service water intake bay level. (Section 1R17)
Inspection Report# : 2007002 (pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : February 04, 2008
 
Indian Point 2 1Q/2008 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS ASSOCIATED WITH AGING CAPACITOR DEGRADATION IN A POWER SUPPLY FOR THE MAIN FEEDWATER SUCTION PRESSURE TRANSMITTER A self-revealing Green finding was identified because Entergy did not implement corrective actions for an adverse condition associated with aging critical power supplies. The inspectors determined that the failure to implement corrective actions was a performance deficiency because it was controry to the requirements of Entergy's procedure EN-LI-102, "Corrective Action Process." Entergy placed this issue in the corrective action program and initiated actions to replace all single-point vulnerable instrument power supplies and all high critical instrument power supplies at both Indian Point Unit 2 and Indian Point Unit 3 that have not already been replaced.
The inspectors determined this finding was more than minor because it wass associated with the Equipment Performance attribute of the Initiang Events cornerstone, and it impacted the cornerstone; and it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems. Specifically, aging capacitors caused the failure of the power supply to the feedwater low suction pressure transmitter, which caused a reduction of main boiler feed pump speeds and resulted in operators initiating a manual reactor trip on February 28, 2007. The inspectors evaluated the significance of this finding usning Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding was determined to be of very low safety significance (Green) because, while it was a transient initiator that resulted in a reactor trip, it did not contribute to the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables.
A self-revealing, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables. As a result, Entergy maintenance personnel installed undersized terminal lugs for the 21 service water pump motor jumper cables on January 26, 2000, which resulted in a high resistance connection that degraded over time and eventually caused the cables to fail while the pump was in service on January 27, 2008. Entergy entered this issue into the corrective action program, replaced the jumper cables with insulated bus bars, tested the motor for damage, and changed Engineering Standard ENN-EE-S-008-IP, IPEC [Indian Point Energy Center] Electrical Cable Installation Standard, to ensure the use of correctly-sized terminal lugs in the future. [Entergy also plans to perform an extent of condition review that includes thermography and visual inspections of other safety related motor cable terminations.]
The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone; and, it affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to provide adequate procedural steps to ensure that the 21 service water pump was installed with appropriate
 
electrical connectors. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train for greater than its Technical Specification allowed outage time; and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events.
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED FIRE BARRIER IN EDG BUILDING The inspectors identified a Green non-cited violation (NCV) of Unit 2 license condition 2.K. because Entergy failed to identify a degraded fire barrier in the emergency diesel generator (EDG) room. Specifically, the inspectors identified a backflow preventer valve in an EDG sump that could not perform its function due to a large allen wrench that was positioned in a manner that would prevent the valve from shutting. Entergy removed the tool, verified functionality of the valve, and entered this condition into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inability of the backflow prevent alve to perform its function represented "moderate" degradation based on the size of the drain line, and the distance between the EDG sumps. The inspectors determined that this issue was of very low safety significance (Green) because the degradation of the fire barrier was moderate, and there was a non-degraded automatic, water-based fire suppression system in the affected fire area.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel routinely conduct tours in the EDG building and had not identified the degraded condition of the backflow preventer valve. (P.1(a))
Inspection Report# : 2007005 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation Degraded 12 Fire Main Booster Pump Cell Fire Door The inspectors identified a Green non-cited violation (NCV) of License Condition 2.K.,
fire protection program, because Entergy failed to identify a degraded three-hour rated fire door on the east entrance of the 12 fire main booster pump room. The door was determined to be inoperable due to a misalignment, which prevented the door from fully closing. Entergy entered this issue into their corrective action program, took immediate compensatory actions, realigned the door, and ensured that it would fully close.
The inspectors determined that this finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone; and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined that this issue was of very low safety significance because the degradation of the fire barrier was moderate based on the fire door displaying significant degradation affecting its performance or reliability.
However, it was still expected to provide some defense-in-depth benefit. Specifically, the fire door was expected to provide a minimum of 20 minutes fire endurance
 
protection, and the in-situ fire ignition sources and flammable materials were positioned such that the degraded fire door would not be subject to direct flame impingement.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel who routinely traverse through or past the fire door had not identified the degraded condition. (P.1(a))
Inspection Report# : 2007004 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation PROCEDURE INADEQUATE TO ENSURE OPERABILITY OF SI PUMPS DURING VENTING The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not ensure that procedures associated with operation of the safety injection (SI) system during venting were appropriate to the circumstances. Specifically, procedure 2-PT-M108, RHR/SI [residual heat removal/safety injection] System Venting, did not have appropriate controls to ensure the safety injection piping and pumps remained operable during accident conditions. Entergy entered the issue into their corrective action program and revised the venting procedure to ensure operator actions are appropriately evaluated and credited to maintain operability of the system.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone; and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined this finding resulted in a loss of function of a single train of SI for approximately five minutes. Because the total inoperability time was less than the allowed outage time of 72 hours, and the finding is not potentially risk significant due to a seismic, flooding, or severe weather initiating event, this finding screens as very low safety significance (Green).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that complete, accurate and up-to-date procedures were available. (H.2(c))
Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO PREVENT EXCEEDING PM FREQUENCY FOR 25 FCU A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy failed to implement effective corrective actions for a condition adverse to quality associated with reduced flow to the containment fan cooler units due to fouling, which resulted from exceeding the periodicity of preventative maintenance activities to clean and inspect the containment fan cooler units. On September 16, 2007, the 25 containment fan cooler unit was declared inoperable due to inadequate service water flow caused by partial fouling of the heat exchanger. Entergy implemented actions to restore service water flow to the 25 containment fan cooler unit, and they entered this issue into their corrective action program to schedule the maintenance on other containment fan cooler units, and to evaluate the appropriate periodicity for the preventative maintenance activity.
The inspectors determined that this finding was more than minor because it was associated with the Structures, Systems, and Components and Barrier Performance attribute of the Barrier Integrity cornerstone; and, it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to take effective corrective actions to
 
prevent exceeding the periodicity for the cleaning and inspection of the 25 containment fan cooler unit resulted in partial flow blockage to the component, and a reduction in flow below the value required by Technical Specifications.
The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green) because it did not impact a function that was important to large early release frequency.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS FOR DEGRADED CONTAINMENT FAN COOLER UNIT SERVICE WATER FLOW The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy failed to implement corrective actions to monitor a condition adverse to quality associated with degradation of service water flow rates to the fan cooler units following the failure of surveillance test 2-PT-Q016, "Containment fan cooler Unit Cooling Water Flow Test," Revision 1, on September 16, 2007. Entergy's corrective actions, which had been developed following failure of the 25 containment fan cooler unit to pass the surveillance flow acceptance criteria on September 16, 2007, included compensatory measures for operations personnel to monitor service water flow to the containment fan cooler unit and to increase the frequency of the quarterly surveillance test. Operations personnel recorded the five containment fan cooler unit service water flow rates in the unit narrative logs, but did not effectively monitor the service water flow rates. Consequently, Entergy failed to identify degrading service water flow and take action prior to the containment fan cooler units being rendered inoperable due to insufficient flow on October 14, 2007. Entergy entered this issue into the corrective action program and updated their action plan to begin systematic trending of service water flows to the containment fan cooler units until the next refueling outage.
The inspectors determined this finding was more than minor in accordance with IMC 0612, Appendix E, "Examples of Minor Issues," Example 3.g, because the failure to implement a corrective action contributed to the service water flows being out-of-specification to all five containment fan cooler units. The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green), because it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not effectively implement corrective actions for a condition adverse to quality associated with degradation of service water flow to containment fan cooler units. (P.1(d))
Inspection Report# : 2007005 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS TO REPAIR A DEGRADED SERVICE WATER FLOW INSTRUMENT The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, in that, Entergy did not implement timely corrective actions for a degraded condition associated with the 25 Containment Fan Cooler Unit (FCU) flow indicator. Specifically, the failure to take timely corrective actions for the degraded service water flow indicator for the 25 FCU, initially identified in October 2006, resulted in the inability to ensure that sufficient service water flow was available for the component to perform its intended function. Subsequently, it was identified that a reduced service water flow condition did exist. Entergy entered the issue into their corrective action program and implemented corrective actions to restore adequate indication of service water flow to the 25 FCU.
Entergy is evaluating maintenance practices to determine the appropriateness of a periodic blow-down of the transmitter impulse lines to prevent sediment buildup.
 
The inspectors determined that this finding was more than minor because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that the physical design barrier (containment) protects the public from radionuclide releases caused by accidents or events. This finding was evaluated using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance because, while it could impact late containment failure, it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the condition when initially identified. (P.1(c))
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : June 05, 2008
 
Indian Point 2 2Q/2008 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Plant Start-Up Procedure Regarding MBFP Turbine Runback Arm/Defeat Switch A Green, self-revealing non-cited violation (NCV) of Technical Specification 5.4.1, Administrative Controls -
Procedures, was identified, because Entergy did not implement the requirements of plant startup procedure 2-POP-1.3, "Plant Startup from Zero To 45% Power. Specifically, operators performed a step out of sequence in the plant operating procedure that was not warranted by plant conditions, and resulted in a main turbine runback followed by a manual reactor trip initiated by control room operators. Entergy entered this issue into the corrective action program, initiated procedural enhancements, performed a post-trip evaluation, and a root cause evaluation.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using the Phase 1 analysis of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance because it did not contribute to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would be unavailable.
The finding had a cross-cutting aspect in the area of human performance because Entergy staff utilized work practices that did not support effective human error prevention techniques by proceeding in the face of uncertainty and unexpected circumstances, when they prematurely positioned the arm/defeat switch contrary to plant procedures and conditions. (H.4(a))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Camera Controls Procedure Resulting in RFI Induced MBFP Runback and Subsequent Manual Reactor Trip A Green, self-revealing finding was identified because Entergy did not implement procedural requirements to evaluate flash photography in the vicinity of sensitive control cabinets. Specifically, Entergy did not implement procedure EN-NS-214, Camera Controls for Access and Use, and evaluate the potential impact of flash photography on sensitive control circuitry. Radiofrequency interference (RFI) from the digital camera during flash photography resulted in a main boiler feed pump runback which required a subsequent manual reactor trip. Entergy entered the issue into the corrective action process, performed site-wide training regarding the potential impacts of RFI from digital cameras on digital plant equipment and reinforced expectations to site personnel regarding procedural compliance.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
The inspectors determined that this finding has a cross-cutting aspect in the area of human performance because
 
Entergy did not effectively communicate expectations regarding procedural compliance and personnel did not follow the applicable procedures. (H.4(b))
Inspection Report# : 2008003 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS ASSOCIATED WITH AGING CAPACITOR DEGRADATION IN A POWER SUPPLY FOR THE MAIN FEEDWATER SUCTION PRESSURE TRANSMITTER A self-revealing Green finding was identified because Entergy did not implement corrective actions for an adverse condition associated with aging critical power supplies. The inspectors determined that the failure to implement corrective actions was a performance deficiency because it was controry to the requirements of Entergy's procedure EN-LI-102, "Corrective Action Process." Entergy placed this issue in the corrective action program and initiated actions to replace all single-point vulnerable instrument power supplies and all high critical instrument power supplies at both Indian Point Unit 2 and Indian Point Unit 3 that have not already been replaced.
The inspectors determined this finding was more than minor because it wass associated with the Equipment Performance attribute of the Initiang Events cornerstone, and it impacted the cornerstone; and it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems. Specifically, aging capacitors caused the failure of the power supply to the feedwater low suction pressure transmitter, which caused a reduction of main boiler feed pump speeds and resulted in operators initiating a manual reactor trip on February 28, 2007. The inspectors evaluated the significance of this finding usning Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding was determined to be of very low safety significance (Green) because, while it was a transient initiator that resulted in a reactor trip, it did not contribute to the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Site Procurement Procedure for EDG Temperature Control Valve Elements The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings because Entergy personnel did not implement the requirements of procedure SAO-270, Procurement Program, for the procurement of safety related temperature control valve (TCV) elements for the emergency diesel generators (EDGs). Specifically, Entergy did not perform a technical evaluation as required for the TCV elements which resulted in the purchase and installation of incorrect TCV elements on the 21 and 22 EDGs between 2002 and 2003.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because the installation of incorrect TCV elements represented a design deficiency that was confirmed not to result in a loss of operability of the EDGs. Specifically, engineering analysis verified past EDG operability was maintained based on analysis that assumed the highest observed service water temperature over the past three years. Entergy entered this issue into the corrective action program and installed the correct TCV elements in 21 and 22 EDGs.
Inspection Report# : 2008003 (pdf)
 
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Station Blackout/Appendix-R Diesel Generator Post Modification Test Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the requirements of EN-DC-117, Post Modification Testing and Special Instructions, to control revisions to the station blackout/Appendix R diesel generator (SBO/App-R DG) post modification test, or to review and approve the test results. Specifically, the SBO/App-R DG post modification test was not sufficient to demonstrate the SBO/App-R DG could perform its intended design functions. As a corrective measure, Entergy subsequently performed additional testing to demonstrate system operability.
The inspectors determined the finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the post modification test deficiencies represented reasonable doubt regarding the operability of the SBO/App-R DG. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergy's supervisory and management oversight of work activities was not adequate to ensure testing was properly performed. H.4(c))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operating Procedure for Station Blackout/Appendix-R Diesel Generator The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because the SBO/App-R DG operating procedure 2-SOP-27.6, "Appendix-R Diesel Generator Operation," was not adequate. Specifically, the procedure could not be performed as written, and was not sufficient to ensure operators could start the SBO/App-R DG, and energize an electrical bus within the required time of one hour. Entergy subsequently revised the procedure to correct the most critical deficiencies, and pre-staged equipment to reduce the time needed to energize a bus. As an interim corrective measure, Entergy relied upon operator training for other deficiencies, pending final corrective actions.
The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure deficiencies resulted in a reasonable doubt whether the SBO/App-R DG could be started and aligned in a timely and correct manner, as required by design. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergys procedure for the SBO/App-R DG was not adequate to assure nuclear safety in implementing necessary operator actions for a SBO.
(H.2(c))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Design Control Associated with a Temporary Modification to Emergency Diesel Generator
 
Service Water Return Piping The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control because Entergy did not adequately analyze, document, or translate seismic considerations for temporary service water hoses installed on the 21 and 23 emergency diesel generator (EDG) heat exchangers during the March 2008 refueling outage. Entergy entered the issue into the corrective action program, evaluated past operability concerns, and added restraints to the temporary service water hoses.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of the EDG system during a Seismic Class I design basis event. This finding was evaluated using IMC 0609, Appendix G, attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors]. The finding was determined to be of very low safety significance (Green) because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. Entergy performed a subsequent operability evaluation which provided reasonable assurance that the EDGs would have performed the safety function during a design basis seismic event.
The finding had a cross-cutting aspect in the area of human performance because Entergy personnel made non-conservative assumptions regarding the seismic adequacy of the temporary hose modification. Specifically, Entergy personnel did not perform an engineering analysis to validate their assumptions that the temporary service water hoses would not adversely impact the seismic qualification of the EDGs. (H.1(b))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Quality Records for Containment Sump Modification The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, because Entergy did not maintain sufficient records to furnish evidence that a safety-related containment sump modification was performed in accordance with the design documentation. Specifically, nine of 63 work orders completed during the 2R17 refueling outage for the modification were missing data or missing entirely due to being lost, misplaced, or contaminated during implementation of the project. Entergy entered the issue into the corrective action process, evaluated the operability impact of the missing data, and performed visual inspections of accessible safety-related welds during the 2R18 refueling outage.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because the finding did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events initiating events. Entergy performed inspections during 2R18 and completed technical evaluations of missing data that provided reasonable assurance of sump operability.
The finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately coordinate work activities to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination was necessary to assure plant and human performance. (H.3(b))
Inspection Report# : 2008003 (pdf)
Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables.
A self-revealing, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix
 
B, Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables. As a result, Entergy maintenance personnel installed undersized terminal lugs for the 21 service water pump motor jumper cables on January 26, 2000, which resulted in a high resistance connection that degraded over time and eventually caused the cables to fail while the pump was in service on January 27, 2008. Entergy entered this issue into the corrective action program, replaced the jumper cables with insulated bus bars, tested the motor for damage, and changed Engineering Standard ENN-EE-S-008-IP, IPEC [Indian Point Energy Center] Electrical Cable Installation Standard, to ensure the use of correctly-sized terminal lugs in the future. [Entergy also plans to perform an extent of condition review that includes thermography and visual inspections of other safety related motor cable terminations.]
The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone; and, it affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to provide adequate procedural steps to ensure that the 21 service water pump was installed with appropriate electrical connectors. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train for greater than its Technical Specification allowed outage time; and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events.
Inspection Report# : 2008002 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED FIRE BARRIER IN EDG BUILDING The inspectors identified a Green non-cited violation (NCV) of Unit 2 license condition 2.K. because Entergy failed to identify a degraded fire barrier in the emergency diesel generator (EDG) room. Specifically, the inspectors identified a backflow preventer valve in an EDG sump that could not perform its function due to a large allen wrench that was positioned in a manner that would prevent the valve from shutting. Entergy removed the tool, verified functionality of the valve, and entered this condition into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inability of the backflow prevent alve to perform its function represented "moderate" degradation based on the size of the drain line, and the distance between the EDG sumps. The inspectors determined that this issue was of very low safety significance (Green) because the degradation of the fire barrier was moderate, and there was a non-degraded automatic, water-based fire suppression system in the affected fire area.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel routinely conduct tours in the EDG building and had not identified the degraded condition of the backflow preventer valve. (P.1(a))
Inspection Report# : 2007005 (pdf)
Significance:      Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation Degraded 12 Fire Main Booster Pump Cell Fire Door The inspectors identified a Green non-cited violation (NCV) of License Condition 2.K.,
fire protection program, because Entergy failed to identify a degraded three-hour rated fire door on the east entrance of the 12 fire main booster pump room. The door was determined to be inoperable due to a misalignment, which prevented the door from fully
 
closing. Entergy entered this issue into their corrective action program, took immediate compensatory actions, realigned the door, and ensured that it would fully close.
The inspectors determined that this finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone; and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined that this issue was of very low safety significance because the degradation of the fire barrier was moderate based on the fire door displaying significant degradation affecting its performance or reliability.
However, it was still expected to provide some defense-in-depth benefit. Specifically, the fire door was expected to provide a minimum of 20 minutes fire endurance protection, and the in-situ fire ignition sources and flammable materials were positioned such that the degraded fire door would not be subject to direct flame impingement.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel who routinely traverse through or past the fire door had not identified the degraded condition. (P.1(a))
Inspection Report# : 2007004 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation PROCEDURE INADEQUATE TO ENSURE OPERABILITY OF SI PUMPS DURING VENTING The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not ensure that procedures associated with operation of the safety injection (SI) system during venting were appropriate to the circumstances. Specifically, procedure 2-PT-M108, RHR/SI [residual heat removal/safety injection] System Venting, did not have appropriate controls to ensure the safety injection piping and pumps remained operable during accident conditions. Entergy entered the issue into their corrective action program and revised the venting procedure to ensure operator actions are appropriately evaluated and credited to maintain operability of the system.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone; and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined this finding resulted in a loss of function of a single train of SI for approximately five minutes. Because the total inoperability time was less than the allowed outage time of 72 hours, and the finding is not potentially risk significant due to a seismic, flooding, or severe weather initiating event, this finding screens as very low safety significance (Green).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that complete, accurate and up-to-date procedures were available. (H.2(c))
Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO PREVENT EXCEEDING PM FREQUENCY
 
FOR 25 FCU A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy failed to implement effective corrective actions for a condition adverse to quality associated with reduced flow to the containment fan cooler units due to fouling, which resulted from exceeding the periodicity of preventative maintenance activities to clean and inspect the containment fan cooler units. On September 16, 2007, the 25 containment fan cooler unit was declared inoperable due to inadequate service water flow caused by partial fouling of the heat exchanger. Entergy implemented actions to restore service water flow to the 25 containment fan cooler unit, and they entered this issue into their corrective action program to schedule the maintenance on other containment fan cooler units, and to evaluate the appropriate periodicity for the preventative maintenance activity.
The inspectors determined that this finding was more than minor because it was associated with the Structures, Systems, and Components and Barrier Performance attribute of the Barrier Integrity cornerstone; and, it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to take effective corrective actions to prevent exceeding the periodicity for the cleaning and inspection of the 25 containment fan cooler unit resulted in partial flow blockage to the component, and a reduction in flow below the value required by Technical Specifications.
The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green) because it did not impact a function that was important to large early release frequency.
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS FOR DEGRADED CONTAINMENT FAN COOLER UNIT SERVICE WATER FLOW The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy failed to implement corrective actions to monitor a condition adverse to quality associated with degradation of service water flow rates to the fan cooler units following the failure of surveillance test 2-PT-Q016, "Containment fan cooler Unit Cooling Water Flow Test," Revision 1, on September 16, 2007. Entergy's corrective actions, which had been developed following failure of the 25 containment fan cooler unit to pass the surveillance flow acceptance criteria on September 16, 2007, included compensatory measures for operations personnel to monitor service water flow to the containment fan cooler unit and to increase the frequency of the quarterly surveillance test. Operations personnel recorded the five containment fan cooler unit service water flow rates in the unit narrative logs, but did not effectively monitor the service water flow rates. Consequently, Entergy failed to identify degrading service water flow and take action prior to the containment fan cooler units being rendered inoperable due to insufficient flow on October 14, 2007. Entergy entered this issue into the corrective action program and updated their action plan to begin systematic trending of service water flows to the containment fan cooler units until the next refueling outage.
The inspectors determined this finding was more than minor in accordance with IMC 0612, Appendix E, "Examples of Minor Issues," Example 3.g, because the failure to implement a corrective action contributed to the service water flows being out-of-specification to all five containment fan cooler units. The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green), because it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not effectively implement corrective actions for a condition adverse to quality associated with degradation of service water flow to containment fan cooler units. (P.1(d))
Inspection Report# : 2007005 (pdf)
Significance:      Oct 03, 2007
 
Identified By: NRC Item Type: NCV NonCited Violation UNTIMELY CORRECTIVE ACTIONS TO REPAIR A DEGRADED SERVICE WATER FLOW INSTRUMENT The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, in that, Entergy did not implement timely corrective actions for a degraded condition associated with the 25 Containment Fan Cooler Unit (FCU) flow indicator. Specifically, the failure to take timely corrective actions for the degraded service water flow indicator for the 25 FCU, initially identified in October 2006, resulted in the inability to ensure that sufficient service water flow was available for the component to perform its intended function. Subsequently, it was identified that a reduced service water flow condition did exist. Entergy entered the issue into their corrective action program and implemented corrective actions to restore adequate indication of service water flow to the 25 FCU.
Entergy is evaluating maintenance practices to determine the appropriateness of a periodic blow-down of the transmitter impulse lines to prevent sediment buildup.
The inspectors determined that this finding was more than minor because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that the physical design barrier (containment) protects the public from radionuclide releases caused by accidents or events. This finding was evaluated using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance because, while it could impact late containment failure, it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the condition when initially identified. (P.1(c))
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Jun 11, 2008 Identified By: NRC Item Type: FIN Finding 2008 IP2 Biennial Problem Identification and Resolution Inspection
 
Identification and Resolution of Problems The inspectors concluded that Entergy identified, evaluated, and resolved problems. The inspectors verified that Entergy had taken actions to address previous NRC findings. In general, Entergy personnel identified problems and entered them into the corrective action program (CAP) at a low threshold. The inspectors also determined that Entergy properly screened equipment issues for operability and reportability, as well as prioritized and evaluated them commensurate with their safety significance. Evaluations appropriately considered extent of condition, generic issues, and previous occurrences. However, broader issues involving evaluations into substantive cross-cutting issues were not appropriately prioritized and evaluated commensurate with the significance of the issues.
The inspectors determined that corrective actions addressed the identified causes and were generally implemented in a timely manner. Notwithstanding, the inspectors noted several examples of minor conditions involving identification of issues, prioritization and quality of evaluations, and implementation of corrective actions. Entergys audits and self-assessments were thorough and probing. The inspectors concluded that Entergy identified, reviewed, and applied relevant industry operating experience (OE). Based on interviews, observations of plant activities, and reviews of the CAP and the Employees Concerns Program (ECP), the inspectors determined that site personnel were willing to raise safety issues and to document them in the CAP.
While the inspectors recognized Entergy has reassessed and revised their corrective action plans to address the substantive cross-cutting issue in the area of procedure adequacy, the inspectors concluded that minimal progress had been made in implementation of the planned actions. The inspectors also concluded that Entergy had identified corrective actions and were in the early stages of implementation of corrective action plans to resolve the substantive cross-cutting issue in corrective action implementation Inspection Report# : 2008010 (pdf)
Last modified : August 29, 2008
 
Indian Point 2 3Q/2008 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Plant Start-Up Procedure Regarding MBFP Turbine Runback Arm/Defeat Switch A Green, self-revealing non-cited violation (NCV) of Technical Specification 5.4.1, Administrative Controls - Procedures, was identified, because Entergy did not implement the requirements of plant startup procedure 2-POP-1.3, "Plant Startup from Zero To 45% Power.
Specifically, operators performed a step out of sequence in the plant operating procedure that was not warranted by plant conditions, and resulted in a main turbine runback followed by a manual reactor trip initiated by control room operators. Entergy entered this issue into the corrective action program, initiated procedural enhancements, performed a post-trip evaluation, and a root cause evaluation.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using the Phase 1 analysis of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance because it did not contribute to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would be unavailable.
The finding had a cross-cutting aspect in the area of human performance because Entergy staff utilized work practices that did not support effective human error prevention techniques by proceeding in the face of uncertainty and unexpected circumstances, when they prematurely positioned the arm/defeat switch contrary to plant procedures and conditions. (H.4(a))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Camera Controls Procedure Resulting in RFI Induced MBFP Runback and Subsequent Manual Reactor Trip A Green, self-revealing finding was identified because Entergy did not implement procedural requirements to evaluate flash photography in the vicinity of sensitive control cabinets. Specifically, Entergy did not implement procedure EN-NS-214, Camera Controls for Access and Use, and evaluate the potential impact of flash photography on sensitive control circuitry. Radiofrequency interference (RFI) from the digital camera during flash photography resulted in a main boiler feed pump runback which required a subsequent manual reactor trip. Entergy entered the issue into the corrective action process, performed site-wide training regarding the potential impacts of RFI from digital cameras on digital plant equipment and reinforced expectations to site personnel regarding procedural compliance.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
The inspectors determined that this finding has a cross-cutting aspect in the area of human performance because Entergy did not effectively communicate expectations regarding procedural compliance and personnel did not follow the applicable procedures. (H.4(b))
Inspection Report# : 2008003 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT CORRECTIVE ACTIONS ASSOCIATED WITH AGING CAPACITOR DEGRADATION IN A POWER SUPPLY FOR THE MAIN FEEDWATER SUCTION PRESSURE TRANSMITTER A self-revealing Green finding was identified because Entergy did not implement corrective actions for an adverse condition associated with aging critical power supplies. The inspectors determined that the failure to implement corrective actions was a performance deficiency because it was controry to the requirements of Entergy's procedure EN-LI-102, "Corrective Action Process." Entergy placed this issue in the
 
corrective action program and initiated actions to replace all single-point vulnerable instrument power supplies and all high critical instrument power supplies at both Indian Point Unit 2 and Indian Point Unit 3 that have not already been replaced.
The inspectors determined this finding was more than minor because it wass associated with the Equipment Performance attribute of the Initiang Events cornerstone, and it impacted the cornerstone; and it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems. Specifically, aging capacitors caused the failure of the power supply to the feedwater low suction pressure transmitter, which caused a reduction of main boiler feed pump speeds and resulted in operators initiating a manual reactor trip on February 28, 2007. The inspectors evaluated the significance of this finding usning Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding was determined to be of very low safety significance (Green) because, while it was a transient initiator that resulted in a reactor trip, it did not contribute to the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Aug 15, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Internal Recirculation Pumps
*Green. The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of the internal recirculation pump minimum flow rates. Specifically, Entergy did not verify the adequacy of the pump minimum flow rates for sustained operation under low flow rate conditions or for strong-pump to weak-pump interactions which could result in dead-heading the weaker pump during parallel pump operation. Following identification of the issue, Entergy revised the Emergency Operating Procedures (EOP) to not start a second internal recirculation pump during conditions of high head recirculation, submitted a licensee event report (LER) for each generating unit, and entered the issue into the corrective action program.
The finding was determined to be more than minor because it is associated with the design control attribute of the Mitigating Systems (MS)
Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. On Unit 2, the team determined the finding was of very low safety significance because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.
The deficiency was not indicative of current performance because the modification on Unit 2 was performed in May 2000. Therefore, there was no cross-cutting aspect.
Inspection Report# : 2008012 (pdf)
Significance:        Aug 08, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater System Configuration Control Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the Auxiliary Feedwater (AFW) operating procedures required by Regulatory Guide 1.33 Appendix A. Specifically, the inspectors identified an AFW drain valve that was not in the required position and an AFW isolation valve that was in the correct position but was not locked as required. Entergy evaluated the as-found configuration of the valves and determined that the AFW system operability was not impacted. Entergy also performed system alignment verifications of AFW and other safety-related systems as part of an extent-of-condition review.
The inspectors determined the finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events. Specifically, the inspectors determined that the as-found configuration of the identified components did not adversely impact system operability. The finding had a cross-cutting aspect in the area of human performance because operators did not use adequate self and peer checking techniques when shutting an open drain valve or when attaching a locking device to an isolation valve. (H.4(a))
Inspection Report# : 2008004 (pdf)
Significance:        Jul 29, 2008 Identified By: NRC
 
Item Type: NCV NonCited Violation On-line Leak Repairs Made Without Use of Proper Procedures The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, when Entergy did not implement on-line leak repair procedures to repair a steam leak on valve MS-2A. Specifically, Entergy performed multiple leak sealant injections on valve MS-2A without engineering controls described in station on-line leak repair procedures. Corrective actions planned included reviewing this issue with the planning and component engineering departments and determining if training on the on-line leak sealing procedures is warranted.
The finding was more than minor because, if left uncorrected, inadequate control of leak-sealant injections would become a more significant safety concern. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function. Specifically, Entergys operability evaluation concluded that the sealant that was injected extruded back out of the leak path and likely did not reach the valves seat or hinge. The finding had a cross cutting aspect related to work control in the area of Human Performance. Entergy personnel did not appropriately plan work activities to conduct online leak repairs on a safety related component. Specifically, Entergy did not identify necessary engineering procedures to adequately perform leak seal repairs on MS-2A during the planning process. These procedures provide necessary limitations, contingencies, and abort criteria. (H.3.(a))
Inspection Report# : 2008004 (pdf)
Significance:        Jul 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation City Water Tank Below Required Level due to Inadequate Design Change Implementation The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, because Entergy did not implement portions of an engineering change package for an alarm setpoint change following modification to the city water tank minimum required water volume calculation. As a result, city water tank level dropped below the minimum water level required by the Technical Requirements Manual.
Corrective actions included updating plant procedures and training of personnel.
The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the Cornerstones objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using a phase 1 analysis described in Inspection Manual Chapter 0609 Appendix F, Fire Protection Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the degradation rating was determined to be low. The finding had a cross-cutting aspect related to formally defining the authority and roles for decisions affecting nuclear safety in the area of Human Performance in that Entergy management did not ensure that roles and responsibilities were communicated clearly to a member of the engineering change team responsible for implementing Operations procedure changes. As a result, the proper procedure changes were not made to plant procedures and logs which ultimately led to unmitigated low levels in the city water tank. (H.1(a))
Inspection Report# : 2008004 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Site Procurement Procedure for EDG Temperature Control Valve Elements The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings because Entergy personnel did not implement the requirements of procedure SAO-270, Procurement Program, for the procurement of safety related temperature control valve (TCV) elements for the emergency diesel generators (EDGs). Specifically, Entergy did not perform a technical evaluation as required for the TCV elements which resulted in the purchase and installation of incorrect TCV elements on the 21 and 22 EDGs between 2002 and 2003.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because the installation of incorrect TCV elements represented a design deficiency that was confirmed not to result in a loss of operability of the EDGs. Specifically, engineering analysis verified past EDG operability was maintained based on analysis that assumed the highest observed service water temperature over the past three years. Entergy entered this issue into the corrective action program and installed the correct TCV elements in 21 and 22 EDGs.
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Station Blackout/Appendix-R Diesel Generator Post Modification Test Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not
 
implement the requirements of EN-DC-117, Post Modification Testing and Special Instructions, to control revisions to the station blackout/Appendix R diesel generator (SBO/App-R DG) post modification test, or to review and approve the test results. Specifically, the SBO/App-R DG post modification test was not sufficient to demonstrate the SBO/App-R DG could perform its intended design functions. As a corrective measure, Entergy subsequently performed additional testing to demonstrate system operability.
The inspectors determined the finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the post modification test deficiencies represented reasonable doubt regarding the operability of the SBO/App-R DG. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergy's supervisory and management oversight of work activities was not adequate to ensure testing was properly performed. H.4(c))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operating Procedure for Station Blackout/Appendix-R Diesel Generator The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because the SBO/App-R DG operating procedure 2-SOP-27.6, "Appendix-R Diesel Generator Operation," was not adequate. Specifically, the procedure could not be performed as written, and was not sufficient to ensure operators could start the SBO/App-R DG, and energize an electrical bus within the required time of one hour. Entergy subsequently revised the procedure to correct the most critical deficiencies, and pre-staged equipment to reduce the time needed to energize a bus. As an interim corrective measure, Entergy relied upon operator training for other deficiencies, pending final corrective actions.
The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure deficiencies resulted in a reasonable doubt whether the SBO/App-R DG could be started and aligned in a timely and correct manner, as required by design. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergys procedure for the SBO/App-R DG was not adequate to assure nuclear safety in implementing necessary operator actions for a SBO. (H.2(c))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Design Control Associated with a Temporary Modification to Emergency Diesel Generator Service Water Return Piping The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control because Entergy did not adequately analyze, document, or translate seismic considerations for temporary service water hoses installed on the 21 and 23 emergency diesel generator (EDG) heat exchangers during the March 2008 refueling outage. Entergy entered the issue into the corrective action program, evaluated past operability concerns, and added restraints to the temporary service water hoses.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of the EDG system during a Seismic Class I design basis event. This finding was evaluated using IMC 0609, Appendix G, attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors]. The finding was determined to be of very low safety significance (Green) because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. Entergy performed a subsequent operability evaluation which provided reasonable assurance that the EDGs would have performed the safety function during a design basis seismic event.
The finding had a cross-cutting aspect in the area of human performance because Entergy personnel made non-conservative assumptions regarding the seismic adequacy of the temporary hose modification. Specifically, Entergy personnel did not perform an engineering analysis to validate their assumptions that the temporary service water hoses would not adversely impact the seismic qualification of the EDGs. (H.1
 
(b))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Quality Records for Containment Sump Modification The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, because Entergy did not maintain sufficient records to furnish evidence that a safety-related containment sump modification was performed in accordance with the design documentation. Specifically, nine of 63 work orders completed during the 2R17 refueling outage for the modification were missing data or missing entirely due to being lost, misplaced, or contaminated during implementation of the project. Entergy entered the issue into the corrective action process, evaluated the operability impact of the missing data, and performed visual inspections of accessible safety-related welds during the 2R18 refueling outage.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because the finding did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events initiating events. Entergy performed inspections during 2R18 and completed technical evaluations of missing data that provided reasonable assurance of sump operability.
The finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately coordinate work activities to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination was necessary to assure plant and human performance. (H.3(b))
Inspection Report# : 2008003 (pdf)
Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables.
A self-revealing, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables. As a result, Entergy maintenance personnel installed undersized terminal lugs for the 21 service water pump motor jumper cables on January 26, 2000, which resulted in a high resistance connection that degraded over time and eventually caused the cables to fail while the pump was in service on January 27, 2008. Entergy entered this issue into the corrective action program, replaced the jumper cables with insulated bus bars, tested the motor for damage, and changed Engineering Standard ENN-EE-S-008-IP, IPEC [Indian Point Energy Center] Electrical Cable Installation Standard, to ensure the use of correctly-sized terminal lugs in the future.
[Entergy also plans to perform an extent of condition review that includes thermography and visual inspections of other safety related motor cable terminations.]
The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone; and, it affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to provide adequate procedural steps to ensure that the 21 service water pump was installed with appropriate electrical connectors. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train for greater than its Technical Specification allowed outage time; and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events.
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEGRADED FIRE BARRIER IN EDG BUILDING The inspectors identified a Green non-cited violation (NCV) of Unit 2 license condition 2.K. because Entergy failed to identify a degraded fire barrier in the emergency diesel generator (EDG) room. Specifically, the inspectors identified a backflow preventer valve in an EDG sump that could not perform its function due to a large allen wrench that was positioned in a manner that would prevent the valve from shutting.
Entergy removed the tool, verified functionality of the valve, and entered this condition into the corrective action program.
The inspectors determined that this finding was more than minor because it was associated with the Protection Against External Factors
 
attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inability of the backflow prevent alve to perform its function represented "moderate" degradation based on the size of the drain line, and the distance between the EDG sumps. The inspectors determined that this issue was of very low safety significance (Green) because the degradation of the fire barrier was moderate, and there was a non-degraded automatic, water-based fire suppression system in the affected fire area.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel routinely conduct tours in the EDG building and had not identified the degraded condition of the backflow preventer valve. (P.1(a))
Inspection Report# : 2007005 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation Degraded 12 Fire Main Booster Pump Cell Fire Door The inspectors identified a Green non-cited violation (NCV) of License Condition 2.K.,
fire protection program, because Entergy failed to identify a degraded three-hour rated fire door on the east entrance of the 12 fire main booster pump room. The door was determined to be inoperable due to a misalignment, which prevented the door from fully closing. Entergy entered this issue into their corrective action program, took immediate compensatory actions, realigned the door, and ensured that it would fully close.
The inspectors determined that this finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone; and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined that this issue was of very low safety significance because the degradation of the fire barrier was moderate based on the fire door displaying significant degradation affecting its performance or reliability.
However, it was still expected to provide some defense-in-depth benefit. Specifically, the fire door was expected to provide a minimum of 20 minutes fire endurance protection, and the in-situ fire ignition sources and flammable materials were positioned such that the degraded fire door would not be subject to direct flame impingement.
The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel who routinely traverse through or past the fire door had not identified the degraded condition. (P.1(a))
Inspection Report# : 2007004 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation PROCEDURE INADEQUATE TO ENSURE OPERABILITY OF SI PUMPS DURING VENTING The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not ensure that procedures associated with operation of the safety injection (SI) system during venting were appropriate to the circumstances. Specifically, procedure 2-PT-M108, RHR/SI [residual heat removal/safety injection] System Venting, did not have appropriate controls to ensure the safety injection piping and pumps remained operable during accident conditions. Entergy entered the issue into their corrective action program and revised the venting procedure to ensure operator actions are appropriately evaluated and credited to maintain operability of the system.
The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone; and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined this finding resulted in a loss of function of a single train of SI for approximately five minutes. Because the total inoperability time was less than the allowed outage time of 72 hours, and the finding is not potentially risk significant due to a seismic, flooding, or severe weather initiating event, this finding screens as very low safety significance (Green).
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that complete, accurate and up-to-date procedures were available. (H.2(c))
 
Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS TO PREVENT EXCEEDING PM FREQUENCY FOR 25 FCU A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy failed to implement effective corrective actions for a condition adverse to quality associated with reduced flow to the containment fan cooler units due to fouling, which resulted from exceeding the periodicity of preventative maintenance activities to clean and inspect the containment fan cooler units. On September 16, 2007, the 25 containment fan cooler unit was declared inoperable due to inadequate service water flow caused by partial fouling of the heat exchanger. Entergy implemented actions to restore service water flow to the 25 containment fan cooler unit, and they entered this issue into their corrective action program to schedule the maintenance on other containment fan cooler units, and to evaluate the appropriate periodicity for the preventative maintenance activity.
The inspectors determined that this finding was more than minor because it was associated with the Structures, Systems, and Components and Barrier Performance attribute of the Barrier Integrity cornerstone; and, it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to take effective corrective actions to prevent exceeding the periodicity for the cleaning and inspection of the 25 containment fan cooler unit resulted in partial flow blockage to the component, and a reduction in flow below the value required by Technical Specifications. The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green) because it did not impact a function that was important to large early release frequency.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT CORRECTIVE ACTIONS FOR DEGRADED CONTAINMENT FAN COOLER UNIT SERVICE WATER FLOW The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy failed to implement corrective actions to monitor a condition adverse to quality associated with degradation of service water flow rates to the fan cooler units following the failure of surveillance test 2-PT-Q016, "Containment fan cooler Unit Cooling Water Flow Test," Revision 1, on September 16, 2007. Entergy's corrective actions, which had been developed following failure of the 25 containment fan cooler unit to pass the surveillance flow acceptance criteria on September 16, 2007, included compensatory measures for operations personnel to monitor service water flow to the containment fan cooler unit and to increase the frequency of the quarterly surveillance test. Operations personnel recorded the five containment fan cooler unit service water flow rates in the unit narrative logs, but did not effectively monitor the service water flow rates.
Consequently, Entergy failed to identify degrading service water flow and take action prior to the containment fan cooler units being rendered inoperable due to insufficient flow on October 14, 2007. Entergy entered this issue into the corrective action program and updated their action plan to begin systematic trending of service water flows to the containment fan cooler units until the next refueling outage.
The inspectors determined this finding was more than minor in accordance with IMC 0612, Appendix E, "Examples of Minor Issues,"
Example 3.g, because the failure to implement a corrective action contributed to the service water flows being out-of-specification to all five containment fan cooler units. The inspectors evaluated this finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity, but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance (Green), because it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding has a cross-cutting aspect in the area of problem identification and resolution because Entergy did not effectively implement corrective actions for a condition adverse to quality associated with degradation of service water flow to containment fan cooler units. (P.1(d))
Inspection Report# : 2007005 (pdf)
Significance:        Oct 03, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
UNTIMELY CORRECTIVE ACTIONS TO REPAIR A DEGRADED SERVICE WATER FLOW INSTRUMENT The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, in that, Entergy did not implement timely corrective actions for a degraded condition associated with the 25 Containment Fan Cooler Unit (FCU) flow indicator.
Specifically, the failure to take timely corrective actions for the degraded service water flow indicator for the 25 FCU, initially identified in October 2006, resulted in the inability to ensure that sufficient service water flow was available for the component to perform its intended function. Subsequently, it was identified that a reduced service water flow condition did exist. Entergy entered the issue into their corrective action program and implemented corrective actions to restore adequate indication of service water flow to the 25 FCU. Entergy is evaluating maintenance practices to determine the appropriateness of a periodic blow-down of the transmitter impulse lines to prevent sediment buildup.
The inspectors determined that this finding was more than minor because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that the physical design barrier (containment) protects the public from radionuclide releases caused by accidents or events. This finding was evaluated using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. This was determined to be a Type B finding because it potentially impacted containment integrity but did not result in the increased likelihood of an initiating event. This finding was determined to be of very low safety significance because, while it could impact late containment failure, it did not impact a function that was important to large early release frequency.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy did not thoroughly evaluate the condition when initially identified. (P.1(c))
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Jun 11, 2008 Identified By: NRC Item Type: FIN Finding 2008 IP2 Biennial Problem Identification and Resolution Inspection Identification and Resolution of Problems The inspectors concluded that Entergy identified, evaluated, and resolved problems. The inspectors verified that Entergy had taken actions to address previous NRC findings. In general, Entergy personnel identified problems and entered them into the corrective action program (CAP) at a low threshold. The inspectors also determined that Entergy properly screened equipment issues for operability and reportability, as well as prioritized and evaluated them commensurate with their safety significance. Evaluations appropriately considered extent of condition, generic issues, and previous occurrences. However, broader issues involving evaluations into substantive cross-cutting issues were not appropriately prioritized and evaluated commensurate with the significance of the issues.
The inspectors determined that corrective actions addressed the identified causes and were generally implemented in a timely manner.
Notwithstanding, the inspectors noted several examples of minor conditions involving identification of issues, prioritization and quality of
 
evaluations, and implementation of corrective actions. Entergys audits and self-assessments were thorough and probing. The inspectors concluded that Entergy identified, reviewed, and applied relevant industry operating experience (OE). Based on interviews, observations of plant activities, and reviews of the CAP and the Employees Concerns Program (ECP), the inspectors determined that site personnel were willing to raise safety issues and to document them in the CAP.
While the inspectors recognized Entergy has reassessed and revised their corrective action plans to address the substantive cross-cutting issue in the area of procedure adequacy, the inspectors concluded that minimal progress had been made in implementation of the planned actions.
The inspectors also concluded that Entergy had identified corrective actions and were in the early stages of implementation of corrective action plans to resolve the substantive cross-cutting issue in corrective action implementation Inspection Report# : 2008010 (pdf)
Last modified : November 26, 2008
 
Indian Point 2 4Q/2008 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Plant Start-Up Procedure Regarding MBFP Turbine Runback Arm/Defeat Switch A Green, self-revealing non-cited violation (NCV) of Technical Specification 5.4.1, Administrative Controls -
Procedures, was identified, because Entergy did not implement the requirements of plant startup procedure 2-POP-1.3, "Plant Startup from Zero To 45% Power. Specifically, operators performed a step out of sequence in the plant operating procedure that was not warranted by plant conditions, and resulted in a main turbine runback followed by a manual reactor trip initiated by control room operators. Entergy entered this issue into the corrective action program, initiated procedural enhancements, performed a post-trip evaluation, and a root cause evaluation.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using the Phase 1 analysis of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance because it did not contribute to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would be unavailable.
The finding had a cross-cutting aspect in the area of human performance because Entergy staff utilized work practices that did not support effective human error prevention techniques by proceeding in the face of uncertainty and unexpected circumstances, when they prematurely positioned the arm/defeat switch contrary to plant procedures and conditions. (H.4(a))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Camera Controls Procedure Resulting in RFI Induced MBFP Runback and Subsequent Manual Reactor Trip A Green, self-revealing finding was identified because Entergy did not implement procedural requirements to evaluate flash photography in the vicinity of sensitive control cabinets. Specifically, Entergy did not implement procedure EN-NS-214, Camera Controls for Access and Use, and evaluate the potential impact of flash photography on sensitive control circuitry. Radiofrequency interference (RFI) from the digital camera during flash photography resulted in a main boiler feed pump runback which required a subsequent manual reactor trip. Entergy entered the issue into the corrective action process, performed site-wide training regarding the potential impacts of RFI from digital cameras on digital plant equipment and reinforced expectations to site personnel regarding procedural compliance.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
 
The inspectors determined that this finding has a cross-cutting aspect in the area of human performance because Entergy did not effectively communicate expectations regarding procedural compliance and personnel did not follow the applicable procedures. (H.4(b))
Inspection Report# : 2008003 (pdf)
Mitigating Systems Significance:      Aug 15, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Internal Recirculation Pumps
*Green. The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of the internal recirculation pump minimum flow rates. Specifically, Entergy did not verify the adequacy of the pump minimum flow rates for sustained operation under low flow rate conditions or for strong-pump to weak-pump interactions which could result in dead-heading the weaker pump during parallel pump operation. Following identification of the issue, Entergy revised the Emergency Operating Procedures (EOP) to not start a second internal recirculation pump during conditions of high head recirculation, submitted a licensee event report (LER) for each generating unit, and entered the issue into the corrective action program.
The finding was determined to be more than minor because it is associated with the design control attribute of the Mitigating Systems (MS) Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. On Unit 2, the team determined the finding was of very low safety significance because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.
The deficiency was not indicative of current performance because the modification on Unit 2 was performed in May 2000. Therefore, there was no cross-cutting aspect.
Inspection Report# : 2008012 (pdf)
Significance:      Aug 08, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater System Configuration Control Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the Auxiliary Feedwater (AFW) operating procedures required by Regulatory Guide 1.33 Appendix A. Specifically, the inspectors identified an AFW drain valve that was not in the required position and an AFW isolation valve that was in the correct position but was not locked as required. Entergy evaluated the as-found configuration of the valves and determined that the AFW system operability was not impacted. Entergy also performed system alignment verifications of AFW and other safety-related systems as part of an extent-of-condition review.
The inspectors determined the finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events. Specifically, the inspectors determined that the as-found configuration of the identified components did not adversely impact system operability. The finding had a cross-cutting aspect in the area of human performance because operators did not use adequate self and peer checking techniques when shutting an open drain
 
valve or when attaching a locking device to an isolation valve. (H.4(a))
Inspection Report# : 2008004 (pdf)
Significance:      Jul 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation On-line Leak Repairs Made Without Use of Proper Procedures The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, when Entergy did not implement on-line leak repair procedures to repair a steam leak on valve MS-2A. Specifically, Entergy performed multiple leak sealant injections on valve MS-2A without engineering controls described in station on-line leak repair procedures. Corrective actions planned included reviewing this issue with the planning and component engineering departments and determining if training on the on-line leak sealing procedures is warranted.
The finding was more than minor because, if left uncorrected, inadequate control of leak-sealant injections would become a more significant safety concern. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function. Specifically, Entergys operability evaluation concluded that the sealant that was injected extruded back out of the leak path and likely did not reach the valves seat or hinge. The finding had a cross cutting aspect related to work control in the area of Human Performance. Entergy personnel did not appropriately plan work activities to conduct online leak repairs on a safety related component. Specifically, Entergy did not identify necessary engineering procedures to adequately perform leak seal repairs on MS-2A during the planning process. These procedures provide necessary limitations, contingencies, and abort criteria. (H.3.(a))
Inspection Report# : 2008004 (pdf)
Significance:      Jul 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation City Water Tank Below Required Level due to Inadequate Design Change Implementation The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, because Entergy did not implement portions of an engineering change package for an alarm setpoint change following modification to the city water tank minimum required water volume calculation. As a result, city water tank level dropped below the minimum water level required by the Technical Requirements Manual. Corrective actions included updating plant procedures and training of personnel.
The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the Cornerstones objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using a phase 1 analysis described in Inspection Manual Chapter 0609 Appendix F, Fire Protection Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the degradation rating was determined to be low. The finding had a cross-cutting aspect related to formally defining the authority and roles for decisions affecting nuclear safety in the area of Human Performance in that Entergy management did not ensure that roles and responsibilities were communicated clearly to a member of the engineering change team responsible for implementing Operations procedure changes. As a result, the proper procedure changes were not made to plant procedures and logs which ultimately led to unmitigated low levels in the city water tank. (H.1(a))
Inspection Report# : 2008004 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Site Procurement Procedure for EDG Temperature Control Valve Elements
 
The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings because Entergy personnel did not implement the requirements of procedure SAO-270, Procurement Program, for the procurement of safety related temperature control valve (TCV) elements for the emergency diesel generators (EDGs). Specifically, Entergy did not perform a technical evaluation as required for the TCV elements which resulted in the purchase and installation of incorrect TCV elements on the 21 and 22 EDGs between 2002 and 2003.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because the installation of incorrect TCV elements represented a design deficiency that was confirmed not to result in a loss of operability of the EDGs. Specifically, engineering analysis verified past EDG operability was maintained based on analysis that assumed the highest observed service water temperature over the past three years. Entergy entered this issue into the corrective action program and installed the correct TCV elements in 21 and 22 EDGs.
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Station Blackout/Appendix-R Diesel Generator Post Modification Test Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the requirements of EN-DC-117, Post Modification Testing and Special Instructions, to control revisions to the station blackout/Appendix R diesel generator (SBO/App-R DG) post modification test, or to review and approve the test results. Specifically, the SBO/App-R DG post modification test was not sufficient to demonstrate the SBO/App-R DG could perform its intended design functions. As a corrective measure, Entergy subsequently performed additional testing to demonstrate system operability.
The inspectors determined the finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the post modification test deficiencies represented reasonable doubt regarding the operability of the SBO/App-R DG. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergy's supervisory and management oversight of work activities was not adequate to ensure testing was properly performed. H.4(c))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operating Procedure for Station Blackout/Appendix-R Diesel Generator The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because the SBO/App-R DG operating procedure 2-SOP-27.6, "Appendix-R Diesel Generator Operation," was not adequate. Specifically, the procedure could not be performed as written, and was not sufficient to ensure operators could start the SBO/App-R DG, and energize an electrical bus within the required time of one hour. Entergy subsequently revised the procedure to correct the most critical deficiencies, and pre-staged equipment to reduce the time needed to energize a bus. As an interim corrective measure, Entergy relied upon operator training for other
 
deficiencies, pending final corrective actions.
The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure deficiencies resulted in a reasonable doubt whether the SBO/App-R DG could be started and aligned in a timely and correct manner, as required by design. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergys procedure for the SBO/App-R DG was not adequate to assure nuclear safety in implementing necessary operator actions for a SBO.
(H.2(c))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Design Control Associated with a Temporary Modification to Emergency Diesel Generator Service Water Return Piping The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control because Entergy did not adequately analyze, document, or translate seismic considerations for temporary service water hoses installed on the 21 and 23 emergency diesel generator (EDG) heat exchangers during the March 2008 refueling outage. Entergy entered the issue into the corrective action program, evaluated past operability concerns, and added restraints to the temporary service water hoses.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of the EDG system during a Seismic Class I design basis event. This finding was evaluated using IMC 0609, Appendix G, attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs [Pressurized Water Reactors] and BWRs [Boiling Water Reactors]. The finding was determined to be of very low safety significance (Green) because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. Entergy performed a subsequent operability evaluation which provided reasonable assurance that the EDGs would have performed the safety function during a design basis seismic event.
The finding had a cross-cutting aspect in the area of human performance because Entergy personnel made non-conservative assumptions regarding the seismic adequacy of the temporary hose modification. Specifically, Entergy personnel did not perform an engineering analysis to validate their assumptions that the temporary service water hoses would not adversely impact the seismic qualification of the EDGs. (H.1(b))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Quality Records for Containment Sump Modification The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, because Entergy did not maintain sufficient records to furnish evidence that a safety-related containment sump modification was performed in accordance with the design documentation. Specifically, nine of 63 work orders completed during the 2R17 refueling outage for the modification were missing data or missing entirely due to being lost, misplaced, or contaminated during implementation of the project. Entergy entered the issue into the corrective
 
action process, evaluated the operability impact of the missing data, and performed visual inspections of accessible safety-related welds during the 2R18 refueling outage.
The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because the finding did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events initiating events. Entergy performed inspections during 2R18 and completed technical evaluations of missing data that provided reasonable assurance of sump operability.
The finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately coordinate work activities to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination was necessary to assure plant and human performance. (H.3(b))
Inspection Report# : 2008003 (pdf)
Significance:      Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables.
A self-revealing, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables. As a result, Entergy maintenance personnel installed undersized terminal lugs for the 21 service water pump motor jumper cables on January 26, 2000, which resulted in a high resistance connection that degraded over time and eventually caused the cables to fail while the pump was in service on January 27, 2008. Entergy entered this issue into the corrective action program, replaced the jumper cables with insulated bus bars, tested the motor for damage, and changed Engineering Standard ENN-EE-S-008-IP, IPEC [Indian Point Energy Center] Electrical Cable Installation Standard, to ensure the use of correctly-sized terminal lugs in the future. [Entergy also plans to perform an extent of condition review that includes thermography and visual inspections of other safety related motor cable terminations.]
The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone; and, it affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to provide adequate procedural steps to ensure that the 21 service water pump was installed with appropriate electrical connectors. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train for greater than its Technical Specification allowed outage time; and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events.
Inspection Report# : 2008002 (pdf)
Barrier Integrity Emergency Preparedness
 
Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: SL-IV Dec 31, 2008 Identified By: NRC Item Type: VIO Violation Site Access Procedure Violation Site Access Procedure Violation - SLIV (involved willfulness)
There was no cross-cutting aspect Inspection Report# : 2008014 (pdf)
Significance: N/A Jun 11, 2008 Identified By: NRC Item Type: FIN Finding 2008 IP2 Biennial Problem Identification and Resolution Inspection Identification and Resolution of Problems The inspectors concluded that Entergy identified, evaluated, and resolved problems. The inspectors verified that Entergy had taken actions to address previous NRC findings. In general, Entergy personnel identified problems and entered them into the corrective action program (CAP) at a low threshold. The inspectors also determined that Entergy properly screened equipment issues for operability and reportability, as well as prioritized and evaluated them commensurate with their safety significance. Evaluations appropriately considered extent of condition, generic issues, and previous occurrences. However, broader issues involving evaluations into substantive cross-cutting issues were not appropriately prioritized and evaluated commensurate with the significance of the issues.
The inspectors determined that corrective actions addressed the identified causes and were generally implemented in a timely manner. Notwithstanding, the inspectors noted several examples of minor conditions involving identification of issues, prioritization and quality of evaluations, and implementation of corrective actions. Entergys audits and self-assessments were thorough and probing. The inspectors concluded that Entergy identified, reviewed, and applied relevant industry operating experience (OE). Based on interviews, observations of plant activities, and reviews of the CAP and the Employees Concerns Program (ECP), the inspectors determined that site personnel were willing to raise safety issues and to document them in the CAP.
While the inspectors recognized Entergy has reassessed and revised their corrective action plans to address the substantive cross-cutting issue in the area of procedure adequacy, the inspectors concluded that minimal progress had been made in implementation of the planned actions. The inspectors also concluded that Entergy had identified
 
corrective actions and were in the early stages of implementation of corrective action plans to resolve the substantive cross-cutting issue in corrective action implementation Inspection Report# : 2008010 (pdf)
Last modified : April 07, 2009
 
Indian Point 2 1Q/2009 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Charred Components in EDG Ventilation Motor Control Center #2 The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not identify that an electrical fault occurred in a safety-related motor control center (MCC) prior to re-energizing the MCC. In addition, the damaged portion of the MCC remained energized for 14 days after it was identified. Entergy entered the issue into the corrective action program, trained all operations personnel on the requirements to replace fuses and re-energize electrical equipment, and is revising the operations procedure for operating electrical equipment.
This issue was more than minor because the finding was associated with the external factors attribute of the initiating events cornerstone and impacted the initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power operations.
Specifically, energizing charred electrical components and maintaining them energized for 14 days increased the likelihood of a fire in the EDG building. The condition was evaluated by a Senior Reactor Analyst utilizing phase two of IMC 0609 Appendix F, Fire Protection Significance Determination Process. It was determined that in the event of an electrical fault-induced fire consuming the MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green. The inspectors determined that a cross-cutting aspect was associated with this finding in the area of human performance, in the component of decision making, in the aspect of conservative assumptions.
Specifically, the decision by Operations personnel to replace fuses and re-energize the EDG building MCC without determining the source of acrid odor in the building was non-conservative. The decision by operations to attempt to locally reclose switches on the affected MCC without performing internal visual inspections was non-conservative.
And lastly, after the charred bucket was identified on January 28, 2009, the organizational decision to leave the bucket energized was non-conservative. H.1(b)
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure for EDG Ventilation Motor Control Center #2 The inspectors identified a Green non-cited violation of TS 5.4.1, Procedures, because Entergy did not maintain an adequate maintenance procedure for a safety related motor control center (MCC). Specifically, the eight-year MCC maintenance procedure, which was performed for the first time on April 6, 2008, did not contain an adequate method to identify high resistance connections within the cubicle. Subsequently, a high resistance connection within the MCC developed into a phase-to-phase electrical fault on January 28, 2009. Entergy entered the issue into the corrective action program, scoped the affected MCC and 21 additional MCCs into the sites thermography program, and planned to revise the maintenance procedure.
The inspectors determined that the finding was more than minor because it was associated with the external factors attribute of the initiating events cornerstone and impacted the initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power operations. Specifically, the inadequate maintenance procedure resulted in a phase-to-phase fault on January 28, 2009 which increased the likelihood of a fire in the EDG building. The condition was evaluated by a Senior
 
Reactor Analyst utilizing phase two of IMC 0609 Appendix F, Fire Protection Significance Determination Process. It was determined that in the event of a fire consuming the MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green. The inspectors determined that the finding had a cross-cutting aspect associated with the area of problem identification and resolution, in the component of operating experience (OE), in the aspect of implementation. Specifically, Entergy did not implement industry recommended practices, or an alternate equivalent method, for identifying high resistance connections in electrical switchgear. P.2(b)
Inspection Report# : 2009002 (pdf)
Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Plant Start-Up Procedure Regarding MBFP Turbine Runback Arm/Defeat Switch A Green, self-revealing non-cited violation (NCV) of Technical Specification 5.4.1, Administrative Controls -
Procedures, was identified, because Entergy did not implement the requirements of plant startup procedure 2-POP-1.3, "Plant Startup from Zero To 45% Power. Specifically, operators performed a step out of sequence in the plant operating procedure that was not warranted by plant conditions, and resulted in a main turbine runback followed by a manual reactor trip initiated by control room operators. Entergy entered this issue into the corrective action program, initiated procedural enhancements, performed a post-trip evaluation, and a root cause evaluation.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using the Phase 1 analysis of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance because it did not contribute to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would be unavailable.
The finding had a cross-cutting aspect in the area of human performance because Entergy staff utilized work practices that did not support effective human error prevention techniques by proceeding in the face of uncertainty and unexpected circumstances, when they prematurely positioned the arm/defeat switch contrary to plant procedures and conditions. (H.4(a))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Camera Controls Procedure Resulting in RFI Induced MBFP Runback and Subsequent Manual Reactor Trip A Green, self-revealing finding was identified because Entergy did not implement procedural requirements to evaluate flash photography in the vicinity of sensitive control cabinets. Specifically, Entergy did not implement procedure EN-NS-214, Camera Controls for Access and Use, and evaluate the potential impact of flash photography on sensitive control circuitry. Radiofrequency interference (RFI) from the digital camera during flash photography resulted in a main boiler feed pump runback which required a subsequent manual reactor trip. Entergy entered the issue into the corrective action process, performed site-wide training regarding the potential impacts of RFI from digital cameras on digital plant equipment and reinforced expectations to site personnel regarding procedural compliance.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because it did not contribute to both the likelihood of a
 
reactor trip and the likelihood that mitigation equipment or functions would not be available.
The inspectors determined that this finding has a cross-cutting aspect in the area of human performance because Entergy did not effectively communicate expectations regarding procedural compliance and personnel did not follow the applicable procedures. (H.4(b))
Inspection Report# : 2008003 (pdf)
Mitigating Systems Significance:        Mar 31, 2009 Identified By: NRC Item Type: FIN Finding Failure to Identify Stuck Open Louvers in 11 Fire Pump House The inspectors identified a Green Finding because Entergy did not identify stuck-open louvers in a fire protection pump room following a pump test on January 14, 2009. The open louvers resulted in freezing conditions in fire protection piping located in the room and cracked two six-inch header isolation valves on January 17, 2009. Entergy entered the issue into the corrective action program and performed a site-wide extent-of-condition walkdown of louvers.
The inspectors determined that the finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the catastrophic failure of the six-inch valves impacted the reliability of the fire header until the ruptured valves were isolated. This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined the issue was of very low safety significance because the cracked valves were easily isolated and did not pass sufficient water to render the fire header non-functional. Specifically, the inspectors assigned a low degradation to the fire header because the fire pumps were able to maintain pressure in the fire header until the ruptured valves were isolated. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, in the component of work practices, in the aspect of human error prevention techniques, because Entergy personnel proceeded in the face of unexpected circumstances.
Specifically, personnel that routinely tour the 11 fire pump house did not question why the room was much colder than normal. H.4(a)
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Promptly Correct Degraded 480 Volt Switchgear Room Fore Door The inspectors identified a Green non-cited violation of License Condition 2.K., fire protection program, because Entergy did not promptly identify and correct a degraded three-hour rated fire door latch mechanism on the west entrance of the 480 Volt switchgear room. Specifically, inspectors identified the fire door in a non-functional state on February 6, 2009, again on February 18, 2009, and again on March 3, 2009. Entergy replaced the fire door latch mechanism on March 3, 2009. This issue was entered into the corrective action program as six condition reports spanning several weeks and included an extent of condition walkdown on March 3, 2009 of all site fire doors.
The finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in the event of a large fire in the 480-Volt switchgear room or the Turbine Building, the affected fire door is credited to prevent the spread of fire from one area to the other area. When degraded, this door impacts the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon during a large fire in the turbine building, and vice versa. This finding was
 
evaluated using Phase X of Inspection Manual Chapter (IMC) 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel did not thoroughly evaluate a degraded fire door latch on February 6, 2009, again on February 18, 2009, and again on February 23, 2009, such that the resolution addressed the cause. P.1(c)
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Include RWST Level Indicator Maintenance in Online Risk Assessment The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(4), because Entergy did not adequately assess and manage the risk associated with the unavailability of the Refueling Water Storage Tank (RWST) level indication during planned maintenance on the level transmitters and instrumentation. Entergy entered the issue into the corrective action program, updated the risk model to include the maintenance activity, assessed the risk, and appropriately coded the maintenance activity to ensure it would be risk assessed in the future.
The inspectors determined that this finding was more than minor because Entergy failed to consider risk significant SSCs that were unavailable during maintenance. The RWST level indication is specifically listed in Table 2 of the plant specific Phase 2 SDP risk-informed inspection notebook. The inspectors determined the significance of this issue in accordance with Inspection Manual Chapter (IMC) 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The inspectors determined that this finding was of very low safety significance because the Incremental Core Damage Probability Deficit was less than 1E-6. The inspectors determined that the finding had a cross-cutting aspect related to using risk insights to plan work in the Work Control component of the Human Performance area. Specifically, Entergy did not appropriately plan work activities by incorporating risk insights for affected plant equipment. H.3(a)
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Acceptance Criteria for Auxiliary Component Cooling Check Valves The inspectors identified a non-cited violation of 10 CFR 50.55a, Codes and standards, because Entergys procedure, 2-PT-Q031A, did not contain appropriate acceptance criteria for positively determining that safety related check valves performed their safety function when required in accordance with the American Society of Mechanical Engineers (ASME) OM Code. Specifically, the test used reverse rotation of a parallel pump to verify that the pumps discharge check valve was closed although past performance demonstrated that the pump, in fact, does not spin backwards when the check valve was stuck open. Entergy entered this issue into their corrective action program as CR-2009-1312.
The inspectors determined that the performance deficiency was greater than minor because it was associated with the Procedure Quality attribute of the Mitigating System cornerstone and it adversely affected the cornerstones objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve 755A reliably performed its safety function when tested as demonstrated by testing performed in January 2005. The inspectors determined that the performance deficiency was of very low safety significance (Green) using NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Specifically, the inspectors answered no to all of the questions in the Mitigating Systems Cornerstone column of the characterization worksheet. The performance deficiency had a cross-cutting aspect related to appropriate corrective actions in the Corrective Action Program component of the Problem Identification and Resolution area. P.1(d).
Inspection Report# : 2009002 (pdf)
 
Significance:      Aug 15, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Internal Recirculation Pumps
*Green. The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of the internal recirculation pump minimum flow rates. Specifically, Entergy did not verify the adequacy of the pump minimum flow rates for sustained operation under low flow rate conditions or for strong-pump to weak-pump interactions which could result in dead-heading the weaker pump during parallel pump operation. Following identification of the issue, Entergy revised the Emergency Operating Procedures (EOP) to not start a second internal recirculation pump during conditions of high head recirculation, submitted a licensee event report (LER) for each generating unit, and entered the issue into the corrective action program.
The finding was determined to be more than minor because it is associated with the design control attribute of the Mitigating Systems (MS) Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. On Unit 2, the team determined the finding was of very low safety significance because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.
The deficiency was not indicative of current performance because the modification on Unit 2 was performed in May 2000. Therefore, there was no cross-cutting aspect.
Inspection Report# : 2008012 (pdf)
Significance:      Aug 08, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater System Configuration Control Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the Auxiliary Feedwater (AFW) operating procedures required by Regulatory Guide 1.33 Appendix A. Specifically, the inspectors identified an AFW drain valve that was not in the required position and an AFW isolation valve that was in the correct position but was not locked as required. Entergy evaluated the as-found configuration of the valves and determined that the AFW system operability was not impacted. Entergy also performed system alignment verifications of AFW and other safety-related systems as part of an extent-of-condition review.
The inspectors determined the finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events. Specifically, the inspectors determined that the as-found configuration of the identified components did not adversely impact system operability. The finding had a cross-cutting aspect in the area of human performance because operators did not use adequate self and peer checking techniques when shutting an open drain valve or when attaching a locking device to an isolation valve. (H.4(a))
Inspection Report# : 2008004 (pdf)
Significance:      Jul 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation On-line Leak Repairs Made Without Use of Proper Procedures The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, when Entergy did not
 
implement on-line leak repair procedures to repair a steam leak on valve MS-2A. Specifically, Entergy performed multiple leak sealant injections on valve MS-2A without engineering controls described in station on-line leak repair procedures. Corrective actions planned included reviewing this issue with the planning and component engineering departments and determining if training on the on-line leak sealing procedures is warranted.
The finding was more than minor because, if left uncorrected, inadequate control of leak-sealant injections would become a more significant safety concern. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function. Specifically, Entergys operability evaluation concluded that the sealant that was injected extruded back out of the leak path and likely did not reach the valves seat or hinge. The finding had a cross cutting aspect related to work control in the area of Human Performance. Entergy personnel did not appropriately plan work activities to conduct online leak repairs on a safety related component. Specifically, Entergy did not identify necessary engineering procedures to adequately perform leak seal repairs on MS-2A during the planning process. These procedures provide necessary limitations, contingencies, and abort criteria. (H.3.(a))
Inspection Report# : 2008004 (pdf)
Significance:        Jul 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation City Water Tank Below Required Level due to Inadequate Design Change Implementation The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, because Entergy did not implement portions of an engineering change package for an alarm setpoint change following modification to the city water tank minimum required water volume calculation. As a result, city water tank level dropped below the minimum water level required by the Technical Requirements Manual. Corrective actions included updating plant procedures and training of personnel.
The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the Cornerstones objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using a phase 1 analysis described in Inspection Manual Chapter 0609 Appendix F, Fire Protection Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the degradation rating was determined to be low. The finding had a cross-cutting aspect related to formally defining the authority and roles for decisions affecting nuclear safety in the area of Human Performance in that Entergy management did not ensure that roles and responsibilities were communicated clearly to a member of the engineering change team responsible for implementing Operations procedure changes. As a result, the proper procedure changes were not made to plant procedures and logs which ultimately led to unmitigated low levels in the city water tank. (H.1(a))
Inspection Report# : 2008004 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Site Procurement Procedure for EDG Temperature Control Valve Elements The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings because Entergy personnel did not implement the requirements of procedure SAO-270, Procurement Program, for the procurement of safety related temperature control valve (TCV) elements for the emergency diesel generators (EDGs). Specifically, Entergy did not perform a technical evaluation as required for the TCV elements which resulted in the purchase and installation of incorrect TCV elements on the 21 and 22 EDGs between 2002 and 2003.
The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable
 
consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because the installation of incorrect TCV elements represented a design deficiency that was confirmed not to result in a loss of operability of the EDGs. Specifically, engineering analysis verified past EDG operability was maintained based on analysis that assumed the highest observed service water temperature over the past three years. Entergy entered this issue into the corrective action program and installed the correct TCV elements in 21 and 22 EDGs.
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Station Blackout/Appendix-R Diesel Generator Post Modification Test Deficiencies The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the requirements of EN-DC-117, Post Modification Testing and Special Instructions, to control revisions to the station blackout/Appendix R diesel generator (SBO/App-R DG) post modification test, or to review and approve the test results. Specifically, the SBO/App-R DG post modification test was not sufficient to demonstrate the SBO/App-R DG could perform its intended design functions. As a corrective measure, Entergy subsequently performed additional testing to demonstrate system operability.
The inspectors determined the finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the post modification test deficiencies represented reasonable doubt regarding the operability of the SBO/App-R DG. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergy's supervisory and management oversight of work activities was not adequate to ensure testing was properly performed. H.4(c))
Inspection Report# : 2008003 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operating Procedure for Station Blackout/Appendix-R Diesel Generator The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because the SBO/App-R DG operating procedure 2-SOP-27.6, "Appendix-R Diesel Generator Operation," was not adequate. Specifically, the procedure could not be performed as written, and was not sufficient to ensure operators could start the SBO/App-R DG, and energize an electrical bus within the required time of one hour. Entergy subsequently revised the procedure to correct the most critical deficiencies, and pre-staged equipment to reduce the time needed to energize a bus. As an interim corrective measure, Entergy relied upon operator training for other deficiencies, pending final corrective actions.
The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure deficiencies resulted in a reasonable doubt whether the SBO/App-R DG could be started and aligned in a timely and correct manner, as required by design. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially
 
risk significant due to external events.
The finding had a cross-cutting aspect in the area of human performance because Entergys procedure for the SBO/App-R DG was not adequate to assure nuclear safety in implementing necessary operator actions for a SBO.
(H.2(c))
Inspection Report# : 2008003 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC}}

Latest revision as of 13:53, 29 November 2024

2017 Q1-Q4 ROP Inspection Findings
ML20311A534
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 11/06/2017
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Office of Nuclear Reactor Regulation
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