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{{#Wiki_filter:1Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 1 of 7 Waterford 3 Initiating Events Significance:          Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
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Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                Page 2 of 7 assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                Page 3 of 7 Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 5 of 7 acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
 
1Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : April 01, 2002
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 1 of 7 Waterford 3 Initiating Events Significance:          Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                Page 2 of 7 assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                Page 3 of 7 Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 5 of 7 acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety
 
2Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Public Radiation Safety Significance:            Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:            May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
 
2Q/2000 Inspection Findings - Waterford 3                                                                                              Page 7 of 7 Inspection Report# : 2000006(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Last modified : April 01, 2002
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                Page 1 of 7 Waterford 3 Initiating Events Significance:        Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Waterford 3                                                                                              Page 2 of 7 Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 3 of 7 The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:          Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                Page 5 of 7 Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Occupational Radiation Safety Public Radiation Safety Significance:          Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:          May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the
 
3Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Last modified : March 29, 2002
 
4Q/2000 Inspection Findings - Waterford 3                                                                                              Page 1 of 7 Waterford 3 Initiating Events Significance:        Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                Page 2 of 7 Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 3 of 7 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                Page 5 of 7 this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Occupational Radiation Safety Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found
 
4Q/2000 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Last modified : March 28, 2002
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                Page 1 of 7 Waterford 3 Initiating Events Significance:        Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                Page 2 of 7 The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 3 of 7 but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                Page 5 of 7 remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:        Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Occupational Radiation Safety Public Radiation Safety Significance:          Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:          May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the
 
1Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Last modified : March 28, 2002
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                Page 1 of 7 Waterford 3 Initiating Events Significance:        Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                Page 2 of 7 The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 3 of 7 but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:          Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                Page 5 of 7 remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:        Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Occupational Radiation Safety Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
 
2Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 Inspection Report# : 2001008(pdf)
Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : March 27, 2002
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                Page 1 of 7 Waterford 3 Initiating Events Significance:        Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:        Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 2 of 7 Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:          Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                Page 3 of 7 Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 4 of 7 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 5 of 7 acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 6 of 7 Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
 
3Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 7 of 7 Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : March 26, 2002
 
4Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 1 of 6 Waterford 3 Initiating Events Significance:          Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:          Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping.
This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:          Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
 
4Q/2001 Inspection Findings - Waterford 3                                                                                                Page 2 of 6 Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds.
Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4.
Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated.
These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was
 
4Q/2001 Inspection Findings - Waterford 3                                                                                              Page 3 of 6 of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC
 
4Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 4 of 6 Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action."
This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -
0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 5 of 6 Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance.
Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection
 
4Q/2001 Inspection Findings - Waterford 3                                                                                                  Page 6 of 6 Significance:            May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : March 01, 2002
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 1 of 8 Waterford 3 Initiating Events Significance:          Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:          Mar 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures.
The inspectors identified a violation of Technical Specification 6.8.1 for the failure to perform corrective maintenance on a reactor trip circuit breaker in accordance with established procedures. During installation of a reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. The licensee performed troubleshooting and repair activities on the breaker, and subsequently placed the breaker in service. No record of the troubleshooting or repair activities was made, resulting in an inability to independently verify the specifics of the problem or provide for traceability of parts used, as required by corrective maintenance procedures. This is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, and is in the licensee's corrective action program as Condition Report 2002-0382. The safety significance of this violation was determined to be more than minor because there was a credible impact on safety, by not performing corrective maintenance in accordance with established procedures on safety related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip circuit breakers would have functioned if required.
Inspection Report# : 2001009(pdf)
Significance:          Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet the requirements of the reactivity management program procedure during maintenance work activities The inspectors identified a violation of Technical Specification 6.8.1 for the failure to meet the reactivity management program requirements during the performance of maintenance on Charging Pump A. The work package for charging pump A did not include a completed reactivity management checklist used to document the reactivity management program screening. The reactivity management program requires that work on specified systems such as the charging system be screened for the potential of an inadvertent reactivity change. Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this condition. Additional issues with the reactivity management program were identified. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance activities. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2002-0169 and -0476. This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion and could
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 2 of 8 affect the function of the charging or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change.
Inspection Report# : 2001009(pdf)
Significance:        Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:        Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module. During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:        Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A. This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition.
The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 3 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds. Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC. This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)
(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:        Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4. Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:        Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated. These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:        Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:        Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2002 Inspection Findings - Waterford 3                                                                                  Page 4 of 8 Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:        Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 5 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action." This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve. The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
Inspection Report# : 2000008(pdf)
Significance:        Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -0785. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:        Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself.
Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770.
The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 6 of 8 fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance. Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure.
The inspectors identified a violation of Criterion III of Appendix B to 10 CFR Part 50 for a design change that failed to fully consider the requirements of Article NC-7000, "Protection Against Overpressure," of Section III in the ASME Boiler and Pressure Vessel Code, 1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid. This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure allowed by the ASME Code. This design change also could affect the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2002-0079. This issue was determined to be of very low safety significance because the modification was not installed in the plant and this design did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
Inspection Report# : 2001009(pdf)
Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
 
1Q/2002 Inspection Findings - Waterford 3                                                                                    Page 7 of 8 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-
: 1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the
 
1Q/2002 Inspection Findings - Waterford 3                                                                                  Page 8 of 8 licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate.
Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems. However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : July 22, 2002
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 1 of 12 Waterford 3 Initiating Events Significance:      Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:      Apr 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain proper foreign material exclusion controls The inspectors identified a violation of Technical Specification 6.8.1 for the failure to maintain proper foreign material exclusion controls in accordance with Procedure UNT 007-059 while working to correct a piping misalignment at Check Valve MS-402A, on the Emergency Feedwater AB main steam supply line. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0628. The safety significance of this violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for foreign material in the main steam supply line affecting the ability of the Emergency Feedwater AB turbine to operate as required. This issue was determined to be of very low safety significance because there was no foreign material found in the main steam piping system during the final system closure inspection.
Inspection Report# : 2002002(pdf)
Significance:      Mar 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                        Page 2 of 12 Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures.
The inspectors identified a violation of Technical Specification 6.8.1 for the failure to perform corrective maintenance on a reactor trip circuit breaker in accordance with established procedures. During installation of a reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. The licensee performed troubleshooting and repair activities on the breaker, and subsequently placed the breaker in service. No record of the troubleshooting or repair activities was made, resulting in an inability to independently verify the specifics of the problem or provide for traceability of parts used, as required by corrective maintenance procedures. This is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, and is in the licensee's corrective action program as Condition Report 2002-0382. The safety significance of this violation was determined to be more than minor because there was a credible impact on safety, by not performing corrective maintenance in accordance with established procedures on safety related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip circuit breakers would have functioned if required.
Inspection Report# : 2001009(pdf)
Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet the requirements of the reactivity management program procedure during maintenance work activities The inspectors identified a violation of Technical Specification 6.8.1 for the failure to meet the reactivity management program requirements during the performance of maintenance on Charging Pump A. The work package for charging pump A did not include a completed reactivity management checklist used to document the reactivity management program screening. The reactivity management program requires that work on specified systems such as the charging system be screened for the potential of an inadvertent reactivity change. Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this condition. Additional issues with the reactivity management program were identified. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance activities. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2002-0169 and -0476. This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion and could affect the function of the charging or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change.
Inspection Report# : 2001009(pdf)
Significance:        Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                        Page 3 of 12 bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:      Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module.
During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:      Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A.
This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                    Page 4 of 12 Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds. Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC.
This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:      Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4. Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:      Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated. These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:      Sep 27, 2000 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 5 of 12 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:      Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:      Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process.
The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
Significance:      Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                        Page 6 of 12 Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety.
Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3),
enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action." This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve.
The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 7 of 12 Inspection Report# : 2000008(pdf)
Significance:      Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -0785. The inspectors assessed this issue using the reactor safety significance determination process.
The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:      Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:      Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 8 of 12 inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530.
The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance. Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure.
The inspectors identified a violation of Criterion III of Appendix B to 10 CFR Part 50 for a design change that failed to fully consider the requirements of Article NC-7000, "Protection Against Overpressure," of Section III in the ASME Boiler and Pressure Vessel Code, 1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid. This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure allowed by the ASME Code. This design change also could affect the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded.
This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2002-0079. This issue was determined to be of very low safety significance because the modification was not installed in the plant and this design did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                    Page 9 of 12 spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
Inspection Report# : 2001009(pdf)
Significance:      Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Apr 12, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to survey and control a high radiation area The licensee failed to survey and control a high radiation area correctly. When items containing radioactive material were placed into a trash holding area on the foot elevation of containment on April 7, 2002, the licensee failed to perform a radiation survey, in violation of 10 CFR 20.1501(a), to evaluate the additional hazard and to adjust the placement of the rope barricade and warning signs (posting). Consequently, radiation dose rates at the rope barricade and posting exceeded the dose rates allowed by Technical Specification 6.12 by a factor of two, demonstrating that the rope barricade and posting did not encompass and control the entire high radiation area. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item is documented in the licensee's corrective action program as Condition Report 2002-00726. The finding had a credible impact on safety because workers could have unknowingly worked in the high radiation area outside the barricades and postings. The occurrence involved the potential for individual workers unplanned, unintended doses resulting from actions or conditions contrary to licensee procedures, Technical Specifications or NRC regulations, which could have been significantly greater if people had worked in the area. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had only very low safety significance because it is not an ALARA finding, an overexposure, a situation involving a substantial potential for overexposure, or an item compromising the ability to assess dose.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 10 of 12 Inspection Report# : 2002002(pdf)
Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
Physical Protection Significance:        May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 11 of 12 Miscellaneous Significance:        Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct the cause of voiding in the low pressure safety injection system and to take effective corrective action to preclude repetition of this condition The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct the cause of voiding in the low-pressure safety injection system and to take effective corrective action to preclude repetition of this condition, which had existed for several months prior to the March 23, through April 17,2002, refueling outage (RF11). The licensee failed to correct the root cause of this condition during the refueling outage. A total of seven unplanned entries into Technical Specification 3.5.2 Action (a) were made as a result of voids large enough to render Low-Pressure Safety Injection System Train A or B inoperable during the period April 18 - June 14, 2002. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0818. This violation was determined to have greater than minor safety significance because it had a credible impact on safety and did affect the operability, availability, reliability, and function of the low-pressure safety injection system. This issue was determined to be of very low safety significance because only one train of the low-pressure safety injection system was affected at any one time and the Technical Specification allowed outage time was not exceeded.
Inspection Report# : 2002002(pdf)
Significance:        Dec 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Technical Specification 3.7.3 and exceeding the allowed outage time for Auxiliary Component Cooling Water Pump B The licensee failed to enter the action statement of Technical Specification 3.7.3 and exceeded the allowed outage time of 72 hours to restore Auxiliary Component Cooling Water Pump B to an operable status. This pump was inoperable for approximately 8 days with a wiped outboard pump bearing caused by the licensee's failure to prevent foreign material from entering the pump bearing oil system. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-1399. This violation was determined to have greater than minor safety significance because it had an actual impact on safety and did affect the operability, availability, reliability, and function of Auxiliary Component Cooling Water Pump B. The issue was determined to be of very low safety significance (Green) based on an NRC senior reactor analyst's Significance Determination Process Phase 3 assessment. This item is documented in the licensee's corrective action program as Condition Report 2001-1399.
Inspection Report# : 2002002(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Waterford 3                                                                      Page 12 of 12 issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems.
However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/03/2003
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 1 of 12 Waterford 3 Initiating Events Significance:      Oct 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Inadvertent Reactor Coolant System Pressue Transient With the reactor coolant system in a solid condition, the licensee performed a calibration of the pressurizer pressure wide range channel A instrument. During this calibration, the primary nuclear plant operator observed what he thought to be lowering reactor coolant system pressure based on the instrument being calibrated. He took action to raise pressure which resulted in lifting the low temperature over-pressure protection relief valves which relieved approximately 50 gallons to the containment sump. The operator failed to confirm the apparent pressure condition using other installed instrumentation. A human performance cross-cutting issue was identified involving ineffective communications between control room operators that resulted in the primary nuclear plant operator not being aware of the calibration activity and reliance on a single pressure instrument for pressure control. The inspectors assessed this event using the reactor safety significance determination process. The inspectors found that the event had very low safety significance because the plant systems and components, while challenged, operated as expected and there were multiple sources of reactor coolant system inventory makeup.
Inspection Report# : 2000011(pdf)
Mitigating Systems Significance:      Sep 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Evaluation Resulting in the Failure of the Shutdown Cooling System The licensee failed to adequately address the capability of the shutdown cooling system to perform its safety function after identifying a degraded condition. This resulted in the failure of two shutdown cooling suction isolation valves to open during attempts to line up the plant for shutdown cooling. The associated inadequate operability evaluation was determined to be a violation of Technical Specification 6.8.1(a) and Administrative Procedure LI-102, "Corrective Action Process," Revision 1. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of the shutdown cooling system during shutdown plant conditions. NRC Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process," was utilized to characterize the significance of the issue. During the loss of shutdown cooling on March 23, 2002, multiple systems or components were available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and high-pressure safety injection system. This event did not result in any loss of instrumentation needed for safe shutdown and cooldown of the plant. Based on multiple success paths available for ensuring decay heat removal capability and inventory makeup capability, this event was characterized as having very low safety significance (Section 1R15).
Inspection Report# : 2002003(pdf)
 
3Q/2002 Inspection Findings - Waterford 3                                                                        Page 2 of 12 Significance:        Apr 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain proper foreign material exclusion controls The inspectors identified a violation of Technical Specification 6.8.1 for the failure to maintain proper foreign material exclusion controls in accordance with Procedure UNT 007-059 while working to correct a piping misalignment at Check Valve MS-402A, on the Emergency Feedwater AB main steam supply line. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0628. The safety significance of this violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for foreign material in the main steam supply line affecting the ability of the Emergency Feedwater AB turbine to operate as required. This issue was determined to be of very low safety significance because there was no foreign material found in the main steam piping system during the final system closure inspection.
Inspection Report# : 2002002(pdf)
Significance:        Mar 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures.
The inspectors identified a violation of Technical Specification 6.8.1 for the failure to perform corrective maintenance on a reactor trip circuit breaker in accordance with established procedures. During installation of a reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. The licensee performed troubleshooting and repair activities on the breaker, and subsequently placed the breaker in service. No record of the troubleshooting or repair activities was made, resulting in an inability to independently verify the specifics of the problem or provide for traceability of parts used, as required by corrective maintenance procedures. This is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, and is in the licensee's corrective action program as Condition Report 2002-0382. The safety significance of this violation was determined to be more than minor because there was a credible impact on safety, by not performing corrective maintenance in accordance with established procedures on safety related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip circuit breakers would have functioned if required.
Inspection Report# : 2001009(pdf)
Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet the requirements of the reactivity management program procedure during maintenance work activities The inspectors identified a violation of Technical Specification 6.8.1 for the failure to meet the reactivity management program requirements during the performance of maintenance on Charging Pump A. The work package for charging pump A did not include a completed reactivity management checklist used to document the reactivity management program screening. The reactivity management program requires that work on specified systems such as the charging system be screened for the potential of an inadvertent reactivity change. Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this condition. Additional issues with the reactivity management program were identified. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance activities. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2002-0169 and -0476. This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion
 
3Q/2002 Inspection Findings - Waterford 3                                                                        Page 3 of 12 and could affect the function of the charging or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change.
Inspection Report# : 2001009(pdf)
Significance:      Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to repair deficiencies in Safety Injection Check Valve SI-142A A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Actions), was identified for inadequate corrective measures taken for an issue identified during a previous outage. Low-Pressure Safety Injection Pump A became vapor bound during the performance of a surveillance test due to the presence of nitrogen in the system. The likely source of the gas was identified as nitrogen saturated water from Safety Injection Tank 2B through leaking Safety Injection System Check Valve SI-142A. This valve had exhibited chronic problems and was identified as leaking past its seat prior to Refueling Outage 10 in the Fall of 2000, but repairs were not performed. The violation is more than minor because it had a credible impact on safety. Low-Pressure Safety Injection Pump A became vapor bound during a surveillance test as a result of nitrogen gas in the discharge line. In addition, this condition contributed to voiding in the respective shutdown cooling line. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2001-1295, -1296, and -1348. The finding represents a problem identification and resolution issue where the licensee's corrective actions for Safety Injection System Check Valve SI-142A were not adequate to prevent a nitrogen void formation in Low-Pressure Coolant Injection Train A piping. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance. The Low-Pressure Safety Injection System Train A discharge line void conditions could have existed for a maximum of 9 days and the actual conditions experienced would not have resulted in Low-Pressure Safety Injection Pump A vapor binding while Train A was in the standby condition. No damage to Train A was observed as a result of operating the pump with the discharge piping not completely filled with water. The actual vapor binding of the pump occurred as a result of the train configuration for a surveillance test. Low-Pressure Safety Injection Train B remained unaffected by this event (Section 1R22).
Inspection Report# : 2001007(pdf)
Significance:      Jul 30, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Degraded Chiller Control Circuit due to Inadequate Modification Essential Chiller AB failed to function as required when it automatically tripped on high compressor temperature and high compressor motor temperature. The cause of the failure was identified as a degraded bearing temperature module.
During troubleshooting, it was identified that the module was not properly grounded. Prior to this failure, the chiller had been modified to reroute selected wires to increase chiller reliability. Part of this modification included relocating this ground which resulted in the module degradation and subsequent chiller failure. This was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0900. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the essential chill water system remained available based on essential chiller Trains A and B had not been modified and the system was capable of performing its safety function (Section 1R17).
Inspection Report# : 2001006(pdf)
Significance:      Jul 18, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Postmaintenance Test for Dry Cooling Tower 2 Sump Pump A
 
3Q/2002 Inspection Findings - Waterford 3                                                                    Page 4 of 12 The licensee failed to specify an adequate postmaintenance test for Dry Cooling Tower 2 replacement Sump Pump A.
This pump was replaced under a maintenance action item that stated that the pump required replacement due to a degraded flow condition. The work package did not specify a flow test of the replacement pump to ensure that the originally identified deficiency had been corrected as required by Technical Specification 6.8.1, Appendix A of Regulatory Guide 1.33, Revision 2, and the licensee's Station Administrative Procedure UNT-005-020, "Post Maintenance Testing," Revision 3, Step 5.1.1. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-0819. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the pump was ultimately demonstrated to be operable and a second motor-driven sump pump and a diesel-driven sump pump remained operable and able to perform the safety function of maintaining the dry cooling tower sump and prevent flooding of electrical equipment (Section 1R19).
Inspection Report# : 2001006(pdf)
Significance: N/A Jan 25, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to report condition outside design basis involving main steam isolation valves In July 1998, the licensee failed to report to the NRC the discovery of a condition outside of the design-basis of the plant, as required by 10 CFR 50.73. After correcting errors in previous analyses, the licensee found that the main steam isolation valves (both Trains A and B) may not have closed during an accident within the design-basis specified time of 4.0 seconds. The closure time could have been as high as 6.1 seconds. Although the licensee determined that no safety limits were challenged, the condition exceeded the design-basis of the plant and should have been reported to the NRC.
This was determined to be a violation of 10 CFR 50.73(a)(2)(ii)(B). This nonconforming condition was of low safety significance because new analyses showed that the longer stroke closure time would not have an adverse impact on the results or consequences of all affected accident analyses. Consequently, the violation of 10 CFR 50.73(a)(2)(ii)(B) identified above is categorized at Severity Level IV and is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-382/0013-01) was entered into the licensee's corrective action program as Condition Report 2001-0171.
Inspection Report# : 2000013(pdf)
Significance:      Nov 13, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to have two operable charging pumps prior to entering Mode 4.
Green. On November 13, 2000, the licensee transitioned from Mode 5 to Mode 4 with the control switch for Charging Pump B in the OFF position rather than in the AUTO position as required. Technical Specification 3.1.2.4 required two operable charging pumps prior to entering Mode 4. Technical Specification 3.0.4 specified that entry into an operational mode shall not be made when the conditions for a limiting condition for operation are not met. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency is documented into the licensee's corrective action program as Condition Report 2000-1515. The issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because three the charging pump could have been manually started if required.
Inspection Report# : 2001008(pdf)
Significance:      Sep 28, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate the ratings of 3-hour fire barriers.
The licensee failed to ensure through testing or evaluation that the configurations of Penetration Seals IIIA0204 and IIIA0251 were 3-hour fire rated. These penetration seals separated fire areas containing equipment required for safe shutdown. This was identified as a violation of License Condition 2.C.9, with two examples, and is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this finding
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 5 of 12 into their corrective action program as Condition Report CR-WF3-2000-1153, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequencies were relatively low, and fire detection and suppression systems were not degraded. The licensee subsequently performed a Generic Letter 86-10 evaluation which qualified these penetration seals.
Inspection Report# : 2000007(pdf)
Significance:      Sep 27, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate condition reports for emergency lighting battery test failures.
The licensee failed to initiate corrective action reports to document and evaluate failures of emergency lighting batteries to pass the 8-hour discharge tests. The team determined that five maintenance action items documented emergency lighting batteries that failed their 8-hour discharge tests. However, the failures were not entered into the licensee's corrective action program, as required by procedure. This was identified as a violation of Technical Specification 6.8.1.f. This violation is being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1141 This finding was of very low safety significance because the batteries would have provided lighting for a certain amount of time and handheld lights would be available, if required.
Inspection Report# : 2000007(pdf)
Significance:      Sep 18, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 1-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
In Fire Area RAB-2 (heating and ventilation mechanical room), it was determined that equipment required for safe shutdown of the plant following a fire were not separated by 1-hour fire barriers. Specifically, several cables for the redundant Train A/B of the chilled water system had either missing or damaged 1-hour fire wrap. This was identified as a violation of Operating License Condition 2.C.9, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2000-1088, and the licensee implemented compensatory measures in the affected fire area in accordance with their fire protection program. This finding was of very low safety significance because the ignition frequency was relatively low, fire suppression and detection systems were not degraded, and actions were available to ensure a safe shutdown path in Fire Area RAB-2.
Inspection Report# : 2000007(pdf)
Significance:      Sep 14, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Correct an Out-Of-Tolerance Core Protection Calculator Channel Reactor Trip Condition Green. On September 14, 2000, the licensee identified that the requirements of Technical Specification 3.3.1 for an inoperable Core Protection Calculator Channel B were not met. The data taken during the surveillance indicated that the low departure from nucleate boiling reactor trip signal was out-of-tolerance. The licensee failed to recognize this condition and returned the channel to operable status. This condition had the effect of delaying this trip signal such that it would not have been generated when required. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This deficiency was entered in the licensee's corrective action program as Condition Report 2000-1074. The issue was assessed using the reactor safety significance determination process.
The inspectors found that the issue had very low safety significance because three other core protection calculator channels were operable and capable of generating the required low departure from nucleate boiling reactor trip signal.
Inspection Report# : 2001008(pdf)
 
3Q/2002 Inspection Findings - Waterford 3                                                                        Page 6 of 12 Significance:        Aug 23, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to meet the requirements of Technical Specification 3.3.3.1 The licensee removed Component Cooling Water System Radiation Monitor AB from service to perform maintenance and calibration. With this equipment out of service, Technical Specification 3.3.3.1 requires that samples be taken every 8 hours to detect a potential reactor coolant system to component cooling water system leak at the reactor coolant pump seal water heat exchangers. The licensee entered the technical specification but did not adequately take samples once per 8 hours as required by Action 28. The chosen sample point, allowed by procedure, was located on a dead leg and did not adequately compensate for the inoperable radiation monitor. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0988. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because a subsequent sample showed no abnormal conditions in the component cooling water system and other radiation monitoring instruments in that system were available to detect an abnormal condition although on a delayed basis.
Inspection Report# : 2000010(pdf)
Significance: N/A Aug 01, 2000 Identified By: NRC Item Type: FIN Finding USQ involving automatic resequencing of nonsafety loads to Class 1E bus (Closes URI 9915-01)
During a previous inspection, the NRC inspectors identified an unresolved item involving a potential violation of 10 CFR 50.59 concerning the automatic resequencing of nonsafety loads to the Class 1E bus following a diesel generator start. The Updated Final Safety Analysis Report indicated that nonsafety loads were only reintroduced manually under administrative controls. This issue was determined to be a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety.
Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3),
enforcement discretion was exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation was not issued (EA-99-220). The inspectors found that the issue had very little safety significance because the nonsafety loads had at least single breaker protection and were not ordinarily vulnerable to faulted conditions.
Inspection Report# : 2000008(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate a condition report upon discovery of a condition adverse to quality The inspectors identified during a review of Permanent Plant Modification ER-W3-99-0857-00-00 and previous test records that Shutdown Cooling Header Thermal Relief Valve S-404A failed its bench test and exceeded its design set point by greater than 22 percent on October 6, 1995. The licensee reset Valve SI-405A to within design limits, however, the licensee failed to initiate a condition report for this condition adverse to quality to identify the root cause and apparent condition that may have existed on other relief valves. The failure to initiate a condition report upon discovery of this condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and Site Procedure W2.501, "Corrective Action." This violation is being treated as a Non Cited Violation in accordance with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2000-0822. This issue was characterized as a "green" finding using the significance determination process. It was determined to have a very low risk significance because even though the as-found relief valve pressure set point exceeded its design set point, sufficient margin existed to maintain the integrity of the piping protected by the valve.
The licensee re-set the valve at the time of discovery to its design set point, and the licensee has since tested the valve and found the as-found set point satisfactory.
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 7 of 12 Inspection Report# : 2000008(pdf)
Significance:      Jul 12, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to enter appropriate Technical Specification requirements - three examples Three examples of failure to enter the appropriate Technical Specification Limiting Condition for Operation were identified. These examples included the plant stack wide range gas monitor, Containment Isolation Valve CS-129A, and the fuel handling building crane. The plant stack wide range gas monitor and Valve CS-129A were rendered inoperable to perform maintenance and the fuel handling building crane failed a surveillance test. In each case, the components should have been declared inoperable and the provisions of the applicable Technical Specification should have been entered. The licensee failed to take these actions. Operations Procedure OP-100-014, "Technical Specification and Technical Requirements Compliance," describes the requirements to enter the appropriate Technical Specification action if a component is unable to perform its intended safety function due to surveillance or maintenance. The failure to enter the appropriate Technical Specification actions was a violation of OP-100-014. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0765, -0777, and -0785. The inspectors assessed this issue using the reactor safety significance determination process.
The inspectors found that the issue had very low risk significance because the provisions of the applicable Technical Specification actions were met by default in each case.
Inspection Report# : 2000008(pdf)
Significance:      Jul 10, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate post maintenance testing and ineffective corrective actions for replacement of control switch knobs Three examples of inadequate maintenance were identified for main control board switch knob replacement. The switches were associated with a containment isolation valve, a boric acid makeup pump recirculation valve, and a boric acid makeup pump. The knobs were replaced incorrectly, which introduced a push-to-trip or a push-to-actuate feature that was not in the original design. In addition, the knob replacement activity for the containment isolation valve resulted in damage to the switch assembly itself. Inadequate post maintenance testing failed to identify these conditions. This event is a repeat of two similar events identified in 1999. Corrective actions taken following the 1999 events failed to prevent reoccurrence. The failure to establish effective corrective actions to prevent reoccurrence of improperly installed control switch knobs was a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0770. The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the valves downstream of the containment isolation valve were closed and the boric acid system components would have gone to their safe condition if a safety injection actuation signal is generated.
Inspection Report# : 2000008(pdf)
Significance:      Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure fire extinguishers remained within their current hydrostatic test dates The inspectors identified discrepancies in the portable fire extinguisher monthly inspection process. Discrepancies included inconsistencies between the fire extinguisher list and the corresponding maps of fire extinguisher locations, expired hydrostatic test dates on fire extinguishers, and lack of training for personnel performing the monthly inspections. A total of 35 fire extinguishers with expired or unknown hydrostatic test performance dates were identified. Technical Specification 6.8.1.f, "Fire Protection Program Implementation," required that fire protection procedures shall be implemented. Procedure MM-007-010, "Fire Extinguisher Inspection and Extinguisher Replacement," described the requirements for fire extinguisher inspections. This failure to ensure that fire extinguishers
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 8 of 12 were within their current hydrostatic test date was a violation of Technical Specification 6.8.1.f. This violation is being treated as a noncited violation and is in the corrective action program as Condition Reports 2000-0504 and 2000-0530.
The inspectors assessed this issue using the reactor safety significance determination process. The inspectors found that the issue had very low risk significance because the overall condition of portable fire extinguishers was considered adequate, although degraded.
Inspection Report# : 2000005(pdf)
Significance:        Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish adequate post-maintenance test procedures for Charging Pump AB The inspectors identified that the specified postmaintenance tests conducted following corrective maintenance on Charging Pump AB were not adequate to identify incorrectly performed maintenance. Specifically, inadequate maintenance resulted in oil seals installed incorrectly and low oil pressure. These conditions were not identified during postmaintenance testing and resulted in the equipment being out of service for a longer period of time than was necessary. This failure to establish adequate postmaintenance test procedures was a violation of 10 CFR Part 50, Appendix B, Criterion V. This violation is being treated as a noncited violation and is in the corrective action program as Condition Report 2000-0679. The inspectors assessed this issue using the reactor safety significance determination process. The finding had very low risk significance. Since Charging Pumps A and B were always available, both trains of the chemical and volume control system remained operable.
Inspection Report# : 2000005(pdf)
Barrier Integrity Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions The inspectors identified that the licensee failed to promptly identify and correct a condition adverse to quality, resulting in repetitive failures of solenoid-operated control valves to properly operate. The failure of these valves resulted in loss of the primary containment isolation function for the fire protection system piping penetrating containment. This was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of Valves FP-601A and -
601B to isolate on a containment isolation signal. This finding did not result in an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment. In accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, this issue was characterized as having very low safety significance.
Inspection Report# : 2002003(pdf)
Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure.
The inspectors identified a violation of Criterion III of Appendix B to 10 CFR Part 50 for a design change that failed to fully consider the requirements of Article NC-7000, "Protection Against Overpressure," of Section III in the ASME Boiler and Pressure Vessel Code, 1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 9 of 12 install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid. This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure allowed by the ASME Code. This design change also could affect the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded.
This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2002-0079. This issue was determined to be of very low safety significance because the modification was not installed in the plant and this design did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
Inspection Report# : 2001009(pdf)
Significance:        Jan 28, 2001 Identified By: Licensee Item Type: FIN Finding Resolution of Failed Inside and Outside Containment Isolation Valves The inside and outside containment isolation valves in the primary sampling system failed to stroke to the closed position following completion of a pressurizer degassing operation. Maintenance on both valves had been performed during the last scheduled refueling outage, which introduced a common mode failure mechanism in the same containment penetration. The initial response to these failures was not timely and focused on the valve actuators rather than the actual cause of the failure, which was thermal binding of the valve internals. This issue was entered in the licensee's corrective action program as Condition Report 2001-118. This issue was assessed using the reactor safety significance determination process. The inspectors found that the issue had very low safety significance because the containment penetration was small in diameter (1/2-inch) and the licensee successfully isolated the penetration manually as required by Technical Specifications.
Inspection Report# : 2000013(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Poor Radiological Work Planning During the review of the licensee's Refueling Outage 11 exposure estimates and exposure performance data, the inspectors identified that the Radiation Work Permit 2002-1600, "Health Physics Surveys and Postings," total person-rem exceeded budgeted person-rem by greater than 50 percent (5.7 rem verses 3.5 rem). From a review of the job-in-progress review, the inspectors noted that additional exposure was due, in part, to a higher source term than planned and increased radiation protection support for lower cavity and steam generator work that was not well communicated to the radiation protection staff. Additionally, the licensee did not reevaluate the dose estimate for Radiation Work Permit 2002-1600, when it was known that the actual effective dose rate was higher than planned. The failure to reevaluate and adjust an as low as is reasonably achievable (ALARA) dose estimate was a performance deficiency. The finding was more than minor because it was associated with an Occupational Radiation Safety cornerstone attribute (ALARA Planning) and affected the associated cornerstone objective. Using the Occupational Radiation Safety
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 10 of 12 Significance Determination Process, the inspectors determined the finding to have very low safety significance because actual job dose was more than 5 person-rem, it exceeded the planned intended dose by more than 50 percent, and the station's 3-year rolling average collective dose was less than 135 person-rem.
Inspection Report# : 2002003(pdf)
Significance:        Apr 12, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to survey and control a high radiation area The licensee failed to survey and control a high radiation area correctly. When items containing radioactive material were placed into a trash holding area on the foot elevation of containment on April 7, 2002, the licensee failed to perform a radiation survey, in violation of 10 CFR 20.1501(a), to evaluate the additional hazard and to adjust the placement of the rope barricade and warning signs (posting). Consequently, radiation dose rates at the rope barricade and posting exceeded the dose rates allowed by Technical Specification 6.12 by a factor of two, demonstrating that the rope barricade and posting did not encompass and control the entire high radiation area. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item is documented in the licensee's corrective action program as Condition Report 2002-00726. The finding had a credible impact on safety because workers could have unknowingly worked in the high radiation area outside the barricades and postings. The occurrence involved the potential for individual workers unplanned, unintended doses resulting from actions or conditions contrary to licensee procedures, Technical Specifications or NRC regulations, which could have been significantly greater if people had worked in the area. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had only very low safety significance because it is not an ALARA finding, an overexposure, a situation involving a substantial potential for overexposure, or an item compromising the ability to assess dose.
Inspection Report# : 2002002(pdf)
Public Radiation Safety Significance:        Jun 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Broadleaf control station was not located as described in the Offsite Dose Calculation Manual During NRC Inspection 50-382/99-19, the inspector determined that a portion of the radiological environmental monitoring program was not implemented as described in the Offsite Dose Calculation Manual. Specifically, the broadleaf control station was not located in the least prevalent wind direction, as described. The finding was identified as an unresolved item, pending licensee review of historical information about the sample location. Since that inspection, the licensee had been unable to justify the change in the broadleaf control station location. Technical Specification 6.8.1.j requires that the Radiological Environmental Monitoring Program be implemented as described in the Offsite Dose Calculation Manual. The Offsite Dose Calculation Manual, Attachment 7.23, required that radiological environmental monitoring program be implemented as required by the Technical Requirements Manual, Table 3.12-1. The Technical Requirements Manual , Table 3.12-1 Section 4c, required that the broadleaf control sample point be located in the least prevalent wind direction. The failure to place the broadleaf control station in the least prevalent wind direction is a violation of Technical Specification 6.8.1.j. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the corrective action program as Condition Report 1999-1004. The inspectors assessed this issue using the public radiation safety significance determination process. The inspectors determined that the deficiency had very low risk significance because there was no specific event or abnormal radioactive release associated with the finding. Additionally, had there been an event, the licensee had other radiological environmental monitoring data, so the licensee had maintained the ability to assess the environmental impact.
Inspection Report# : 2000005(pdf)
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 11 of 12 Physical Protection Significance:      May 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadequate protection of Safeguards Information Licensee Event Report 00-S02-00 documented a failure to protect safeguards information. The licensee identified that significant safeguards information had been left on the site local area network for over 3 years. Procedure W5.503, "Handling of Safeguard Information," Revision 7, Section 5.15, requires that safeguards information not be processed, produced, or stored on an automatic data processing system that is connected to a local area or wide area network. This failure was identified as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2000-0524. This issue was assessed using the physical protection significance determination process. The inspectors found that the issue had very low risk significance because there were no similar findings in the last 4 quarters.
Inspection Report# : 2000010(pdf)
Miscellaneous Significance:        Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct the cause of voiding in the low pressure safety injection system and to take effective corrective action to preclude repetition of this condition The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct the cause of voiding in the low-pressure safety injection system and to take effective corrective action to preclude repetition of this condition, which had existed for several months prior to the March 23, through April 17,2002, refueling outage (RF11). The licensee failed to correct the root cause of this condition during the refueling outage. A total of seven unplanned entries into Technical Specification 3.5.2 Action (a) were made as a result of voids large enough to render Low-Pressure Safety Injection System Train A or B inoperable during the period April 18 - June 14, 2002. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0818. This violation was determined to have greater than minor safety significance because it had a credible impact on safety and did affect the operability, availability, reliability, and function of the low-pressure safety injection system. This issue was determined to be of very low safety significance because only one train of the low-pressure safety injection system was affected at any one time and the Technical Specification allowed outage time was not exceeded.
Inspection Report# : 2002002(pdf)
Significance:        Dec 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Technical Specification 3.7.3 and exceeding the allowed outage time for Auxiliary Component Cooling Water Pump B The licensee failed to enter the action statement of Technical Specification 3.7.3 and exceeded the allowed outage time of 72 hours to restore Auxiliary Component Cooling Water Pump B to an operable status. This pump was inoperable for approximately 8 days with a wiped outboard pump bearing caused by the licensee's failure to prevent foreign material from entering the pump bearing oil system. This violation is being treated as a noncited violation consistent
 
3Q/2002 Inspection Findings - Waterford 3                                                                      Page 12 of 12 with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2001-1399. This violation was determined to have greater than minor safety significance because it had an actual impact on safety and did affect the operability, availability, reliability, and function of Auxiliary Component Cooling Water Pump B. The issue was determined to be of very low safety significance (Green) based on an NRC senior reactor analyst's Significance Determination Process Phase 3 assessment. This item is documented in the licensee's corrective action program as Condition Report 2001-1399.
Inspection Report# : 2002002(pdf)
Significance: N/A Jun 22, 2001 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee effectively identified problems and entered them into the corrective action program. This was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee during the review period. The licensee appropriately prioritized, characterized, and evaluated issues that were significant conditions adverse to quality. However, it was noted that human performance was a significant contributor to conditions documented in the corrective action program. The licensee adequately implemented corrective actions commensurate with safety that were generally effective. The licensee acknowledged that effectiveness of corrective actions was an ongoing issue. Licensee audits and assessments critically assessed problem identification and resolution activities and identified needs for improvement, as appropriate. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 30, 2000 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team concluded that the licensee was effective in the identification, resolution, and prevention of problems.
However, the team observed that the licensee's monitoring of equipment deficiencies involving degraded, but operable, components and systems, did not track the corrective actions to completion until recently. Further, the condition review group had not consistently considered the need to address degraded, but operable, conditions of safety-related equipment in prioritizing actions. The licensee identified 57 open condition reports that were not identified in the condition report system as involving degraded, but operable equipment. The team reviewed 5 of these open condition reports and found prioritization of the sample was appropriate and that the licensee had determined that the due dates for completion of corrective actions were responsive. Corrective actions, when specified, were implemented in a timely manner. Licensee audits and assessments were effective in identifying areas of improvement and underlying programmatic problems. Based on the interviews conducted during this inspection, workers at the site felt free to initiate condition reports for safety issues in the licensee's identification and resolution of problems program. The team noted that site personnel clearly understood the importance of this program.
Inspection Report# : 2000006(pdf)
Last modified : December 02, 2002
 
4Q/2002 Inspection Findings - Waterford 3                                                                                                Page 1 of 6 Waterford 3 Initiating Events Mitigating Systems Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow an operating procedure The licensee failed to follow Operating Procedure OP-002-003, "Component Cooling Water System," Revision 13, following maintenance activities on Essential Chiller A. The failure to follow procedure resulted in Component Cooling Water Valve CC-305A being mispositioned on November 22, 2002, affecting operability of both Component Cooling Water System Train A and Essential Chiller AB. The failure to follow an operating procedure is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the mitigating systems objective to ensure the availability and capability of the component cooling water and essential chill water systems were affected. The finding is of very low safety significance since the mispositioned valve did not result in loss of safety function for a single train for greater than the Technical Specification allowed outage time. The condition was promptly identified and corrected by the licensee approximately 1.5 hours after Valve CC-305A was mispositioned.
Inspection Report# : 2002004(pdf)
Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples associated with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b).
Three examples included:
* The licensee failed to promptly identify and correct a degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997. The team found that the licensee failed to evaluate the cause of the oil intrusion until 2001, took no corrective actions in 2001 or 2002 to prevent the oil intrusion when the source was identified, and failed to fully evaluate the detrimental effects that the oil intrusion could pose to the electrical and electronic components. The failure to promptly identify and correct the degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997 was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. 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 since the degraded condition did not result in a loss of the emergency diesel generator safety function.
* The licensee failed to promptly identify and correct a degraded condition resulting in exceeding the rated thermal power limit from February 1995 to March 2002. This condition, identified by the licensee in March 2002, introduced non-conservative excore neutron detector calibration errors which affected the high linear power level, high logarithmic power level, high local power density, and low departure from nucleate boiling ratio, reactor protection trip functions. The failure to promptly identify and correct the overpower condition was determined to be a violation of the facility operating License NPF-38 and 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because it affected four reactor trip functions in a non-conservative manner, thus, potentially impacting the barrier cornerstone integrity. The finding is of very low safety significance since it was determined that the accident analysis, Chapter 15 of the Final Safety Analysis Report, bounded the non-conservative trip functions. This finding is also of very low safety significance since actual fuel barrier integrity was never challenged during the overpower condition.
* On April 18, 2002 when the low pressure safety injection Train B was found voided, the licensee failed to identify that the containment spray system Train B would also be voided from similar plant conditions. The containment spray voiding was identified by the licensee on September 17, 2002, when abnormal indications were noted by operators during a surveillance. Action was then taken by the licensee to correct the degraded condition. However, the licensee failed to identify the degraded condition during previous opportunities. The failure to promptly identify and correct the voided condition affecting containment spray Train B was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected the voided condition could impact the reliability of the containment spray system to perform its safety function during accident conditions. The finding is of very low safety significance since the licensee could demonstrate through analysis that the actual degraded condition found would not have prevented the system from performing its safety function during accident conditions.
Inspection Report# : 2002005(pdf)
 
4Q/2002 Inspection Findings - Waterford 3                                                                                              Page 2 of 6 Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Untimeliness Two examples of failures to implement timely corrective actions to resolve degraded conditions were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.c). Two examples included:
* The licensee failed to promptly identify and correct piping connections susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000. In 1997, the licensee experienced a crack of the charging system header due to fatigue stress cracking and determined additional piping connections were susceptible. The piping connection that failed in March 2000 was identified as being susceptible to fatigue stress cracking, however, no corrective actions had been taken. The failure to promptly identify and correct piping susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000, is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because if left uncorrected the finding could become a more significant event.
The finding is of very low safety significance since the degradation of the system was identified and corrected prior to the safety function of the system being adversely impacted.
* The licensee failed to promptly implement timely corrective actions to operate and maintain the low pressure safety injection system as described in the Final Safety Analysis Report. Specifically, since 1997, the licensee utilized multiple analysis for evaluating degraded piping and pipe supports to evaluate acceptable void sizes. These analysis utilized allowable stresses that exceeded the design criteria allowable stresses described in the facilities Final Safety Analysis Report for the low pressure safety injection system. The failure to implement timely corrective actions to restore and maintain the low pressure safety injection system as described in the Final Safety Analysis Report is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because the Mitigating Systems Objective to ensure the availability, reliability, and capability is potentially affected when the system is maintained outside of its design criteria as described in the Final Safety Analysis Report. The finding is of very low safety significance since the analysis used to assess the degraded condition ensured the system could perform its safety function.
Inspection Report# : 2002005(pdf)
Significance:        Sep 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Evaluation Resulting in the Failure of the Shutdown Cooling System The licensee failed to adequately address the capability of the shutdown cooling system to perform its safety function after identifying a degraded condition. This resulted in the failure of two shutdown cooling suction isolation valves to open during attempts to line up the plant for shutdown cooling. The associated inadequate operability evaluation was determined to be a violation of Technical Specification 6.8.1(a) and Administrative Procedure LI-102, "Corrective Action Process," Revision 1. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of the shutdown cooling system during shutdown plant conditions. NRC Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process," was utilized to characterize the significance of the issue. During the loss of shutdown cooling on March 23, 2002, multiple systems or components were available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and high-pressure safety injection system. This event did not result in any loss of instrumentation needed for safe shutdown and cooldown of the plant. Based on multiple success paths available for ensuring decay heat removal capability and inventory makeup capability, this event was characterized as having very low safety significance (Section 1R15).
Inspection Report# : 2002003(pdf)
Significance:        Apr 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain proper foreign material exclusion controls The inspectors identified a violation of Technical Specification 6.8.1 for the failure to maintain proper foreign material exclusion controls in accordance with Procedure UNT 007-059 while working to correct a piping misalignment at Check Valve MS-402A, on the Emergency Feedwater AB main steam supply line. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0628. The safety significance of this violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for foreign material in the main steam supply line affecting the ability of the Emergency Feedwater AB turbine to operate as required. This issue was determined to be of very low safety significance because there was no foreign material found in the main steam piping system during the final system closure inspection.
Inspection Report# : 2002002(pdf)
 
4Q/2002 Inspection Findings - Waterford 3                                                                                                Page 3 of 6 Significance:        Mar 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures.
The inspectors identified a violation of Technical Specification 6.8.1 for the failure to perform corrective maintenance on a reactor trip circuit breaker in accordance with established procedures. During installation of a reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. The licensee performed troubleshooting and repair activities on the breaker, and subsequently placed the breaker in service. No record of the troubleshooting or repair activities was made, resulting in an inability to independently verify the specifics of the problem or provide for traceability of parts used, as required by corrective maintenance procedures. This is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, and is in the licensee's corrective action program as Condition Report 2002-0382. The safety significance of this violation was determined to be more than minor because there was a credible impact on safety, by not performing corrective maintenance in accordance with established procedures on safety related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip circuit breakers would have functioned if required.
Inspection Report# : 2001009(pdf)
Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet the requirements of the reactivity management program procedure during maintenance work activities The inspectors identified a violation of Technical Specification 6.8.1 for the failure to meet the reactivity management program requirements during the performance of maintenance on Charging Pump A. The work package for charging pump A did not include a completed reactivity management checklist used to document the reactivity management program screening. The reactivity management program requires that work on specified systems such as the charging system be screened for the potential of an inadvertent reactivity change. Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this condition. Additional issues with the reactivity management program were identified. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance activities. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Reports 2002-0169 and -0476. This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion and could affect the function of the charging or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change.
Inspection Report# : 2001009(pdf)
Barrier Integrity Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to establish an adequate maintenance procedure The licensee failed to establish an adequate maintenance procedure to ensure Switchgear Ventilation Damper SVS-102 remained in its safe position during maintenance and after the switchgear ventilation system was returned to an operable condition. Specifically, the damper was worked over a two day period without the damper being gagged in its safety minimum open position. The switchgear ventilation system was returned to an operable condition on September 19, 2002, without the associated actuator having been connected or a gag installed to maintain the damper in the minimal open position. The failure to gag the damper or restore the damper to an operable condition would have prevented the damper from being able to perform its safety function (minimum open position) on a safety injection actuation signal. The failure to provide adequate work instructions to repair Ventilation Damper SVS-102 is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the barrier integrity objective, to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events, was affected. A Phase 3 review was performed that considered the potential impact the switchgear ventilation system could have on the control envelope. The NRC risk analyst considered both radiological and toxic gas atmosphere. This finding is of very low safety significance, in part, based on a redundant damper being operable and the short duration the condition actually existed.
Inspection Report# : 2002004(pdf)
Significance:        Dec 20, 2002
 
4Q/2002 Inspection Findings - Waterford 3                                                                                              Page 4 of 6 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Low Pressure Safety Injection System The licensee failed to maintain design control of the low pressure safety injection system, Train A, in accordance with the design basis, as described in the Final Safety Analysis Report, when installing a modification to mitigate adverse voiding conditions that have affected the system. The failure to maintain design control of the system resulted in loss of a Seismic Class 1, ASME Section III, Safety Class 2, barrier during post accident conditions. The failure to maintain design control of the system is a violation of 10 CFR Part 50, Appendix B, Criterion III. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue screens more than minor because the Barrier Integrity Objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events was potentially affected. The finding is of very low safety significance since only degradation of the radiological barrier function provided for the auxiliary building was affected.
Inspection Report# : 2002005(pdf)
Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions The inspectors identified that the licensee failed to promptly identify and correct a condition adverse to quality, resulting in repetitive failures of solenoid-operated control valves to properly operate. The failure of these valves resulted in loss of the primary containment isolation function for the fire protection system piping penetrating containment. This was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of Valves FP-601A and -601B to isolate on a containment isolation signal. This finding did not result in an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment. In accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, this issue was characterized as having very low safety significance.
Inspection Report# : 2002003(pdf)
Significance:        Jan 18, 2002 Identified By: NRC Item Type: NCV NonCited Violation Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure.
The inspectors identified a violation of Criterion III of Appendix B to 10 CFR Part 50 for a design change that failed to fully consider the requirements of Article NC-7000, "Protection Against Overpressure," of Section III in the ASME Boiler and Pressure Vessel Code, 1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid. This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure allowed by the ASME Code. This design change also could affect the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report CR-WF3-2002-0079. This issue was determined to be of very low safety significance because the modification was not installed in the plant and this design did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
Inspection Report# : 2001009(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jul 19, 2002 Identified By: NRC
 
4Q/2002 Inspection Findings - Waterford 3                                                                                            Page 5 of 6 Item Type: FIN Finding Poor Radiological Work Planning During the review of the licensee's Refueling Outage 11 exposure estimates and exposure performance data, the inspectors identified that the Radiation Work Permit 2002-1600, "Health Physics Surveys and Postings," total person-rem exceeded budgeted person-rem by greater than 50 percent (5.7 rem verses 3.5 rem). From a review of the job-in-progress review, the inspectors noted that additional exposure was due, in part, to a higher source term than planned and increased radiation protection support for lower cavity and steam generator work that was not well communicated to the radiation protection staff. Additionally, the licensee did not reevaluate the dose estimate for Radiation Work Permit 2002-1600, when it was known that the actual effective dose rate was higher than planned. The failure to reevaluate and adjust an as low as is reasonably achievable (ALARA) dose estimate was a performance deficiency. The finding was more than minor because it was associated with an Occupational Radiation Safety cornerstone attribute (ALARA Planning) and affected the associated cornerstone objective. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because actual job dose was more than 5 person-rem, it exceeded the planned intended dose by more than 50 percent, and the station's 3-year rolling average collective dose was less than 135 person-rem.
Inspection Report# : 2002003(pdf)
Significance:        Apr 12, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to survey and control a high radiation area The licensee failed to survey and control a high radiation area correctly. When items containing radioactive material were placed into a trash holding area on the foot elevation of containment on April 7, 2002, the licensee failed to perform a radiation survey, in violation of 10 CFR 20.1501(a), to evaluate the additional hazard and to adjust the placement of the rope barricade and warning signs (posting). Consequently, radiation dose rates at the rope barricade and posting exceeded the dose rates allowed by Technical Specification 6.12 by a factor of two, demonstrating that the rope barricade and posting did not encompass and control the entire high radiation area. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item is documented in the licensee's corrective action program as Condition Report 2002-00726. The finding had a credible impact on safety because workers could have unknowingly worked in the high radiation area outside the barricades and postings. The occurrence involved the potential for individual workers unplanned, unintended doses resulting from actions or conditions contrary to licensee procedures, Technical Specifications or NRC regulations, which could have been significantly greater if people had worked in the area. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had only very low safety significance because it is not an ALARA finding, an overexposure, a situation involving a substantial potential for overexposure, or an item compromising the ability to assess dose.
Inspection Report# : 2002002(pdf)
Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Dec 20, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee's process to identify, prioritize, evaluate, and correct problems was generally effective during calender years 2001 and 2002. The team reviewed 250 condition reports that were opened or closed during the period and found, in general, that station personnel effectively identified, characterized, and prioritized problems. Some issues involving the evaluation and correction of degraded conditions were identified by the team. Most of these issues were associated with longstanding degraded conditions that were identified and corrected by the licensee during this period and included the following: (1) an untimely identification of a void condition in the containment spray system existing between April and September 2002, (2) inadequate extent of condition reviews to identify main steam flow venturi degradation which existed since 1995 and the deleterious affect an oil coating which existed since 1997 would have on electrical components associated with the emergency diesel generator, (3) the inappropriate use of engineering analyses that allowed piping supports to exceed design basis allowable stresses during postulated accidents with voids in the low pressure safety injection system since 1997, (4) an inadequate verification of the design adequacy of a plant modification to vent low pressure safety injection system voids installed in June 2002, and (5) untimely corrective actions which resulted in a forced shutdown to repair weld cracks in the charging system in March 2000. Most of these issues had cross-cutting
 
4Q/2002 Inspection Findings - Waterford 3                                                                                                Page 6 of 6 aspects in the area of problem identification and resolution.
Inspection Report# : 2002005(pdf)
Significance:        Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct the cause of voiding in the low pressure safety injection system and to take effective corrective action to preclude repetition of this condition The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct the cause of voiding in the low-pressure safety injection system and to take effective corrective action to preclude repetition of this condition, which had existed for several months prior to the March 23, through April 17,2002, refueling outage (RF11). The licensee failed to correct the root cause of this condition during the refueling outage. A total of seven unplanned entries into Technical Specification 3.5.2 Action (a) were made as a result of voids large enough to render Low-Pressure Safety Injection System Train A or B inoperable during the period April 18 - June 14, 2002. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0818. This violation was determined to have greater than minor safety significance because it had a credible impact on safety and did affect the operability, availability, reliability, and function of the low-pressure safety injection system. This issue was determined to be of very low safety significance because only one train of the low-pressure safety injection system was affected at any one time and the Technical Specification allowed outage time was not exceeded.
Inspection Report# : 2002002(pdf)
Last modified : March 25, 2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                        Page 1 of 7 Waterford 3 1Q/2003 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2003 Identified By: NRC Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified for the failure to maintain and operate main generator seal oil backup differential pressure regulating Valve SO-308 in accordance with vendor recommendations. This condition resulted in a turbine trip and subsequent reactor power cutback on February 14, 2003. This self-revealing finding is greater than minor because it resulted in a perturbation in plant stability resulting in a reactor power cutback, similar to example 4.b in Appendix E of Manual Chapter 0612. The finding is of very low safety significance because, although it caused a plant transient, it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3).
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow an operating procedure The licensee failed to follow Operating Procedure OP-002-003, "Component Cooling Water System," Revision 13, following maintenance activities on Essential Chiller A. The failure to follow procedure resulted in Component Cooling Water Valve CC-305A being mispositioned on November 22, 2002, affecting operability of both Component Cooling Water System Train A and Essential Chiller AB. The failure to follow an operating procedure is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the mitigating systems objective to ensure the availability and capability of the component cooling water and essential chill water systems were affected. The finding is of very low safety significance since the mispositioned valve did not result in loss of safety function for a single train for greater than the Technical Specification allowed outage time. The condition was promptly identified and corrected by the licensee approximately 1.5 hours after Valve CC-305A was mispositioned.
Inspection Report# : 2002004(pdf)
Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                        Page 2 of 7 Three examples associated (two in Mitigating Systems) with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The mitigating systems examples included:
* The licensee failed to promptly identify and correct a degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997. The team found that the licensee failed to evaluate the cause of the oil intrusion until 2001, took no corrective actions in 2001 or 2002 to prevent the oil intrusion when the source was identified, and failed to fully evaluate the detrimental effects that the oil intrusion could pose to the electrical and electronic components. The failure to promptly identify and correct the degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997 was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. 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 since the degraded condition did not result in a loss of the emergency diesel generator safety function.
* On April 18, 2002 when the low pressure safety injection Train B was found voided, the licensee failed to identify that the containment spray system Train B would also be voided from similar plant conditions. The containment spray voiding was identified by the licensee on September 17, 2002, when abnormal indications were noted by operators during a surveillance. Action was then taken by the licensee to correct the degraded condition. However, the licensee failed to identify the degraded condition during previous opportunities.
The failure to promptly identify and correct the voided condition affecting containment spray Train B was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected the voided condition could impact the reliability of the containment spray system to perform its safety function during accident conditions. The finding is of very low safety significance since the licensee could demonstrate through analysis that the actual degraded condition found would not have prevented the system from performing its safety function during accident conditions.
Inspection Report# : 2002005(pdf)
Significance:      Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Untimeliness Two examples of failures to implement timely corrective actions to resolve degraded conditions were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.c). Two examples included:
* The licensee failed to promptly identify and correct piping connections susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000. In 1997, the licensee experienced a crack of the charging system header due to fatigue stress cracking and determined additional piping connections were susceptible. The piping connection that failed in March 2000 was identified as being susceptible to fatigue stress cracking, however, no corrective actions had been taken. The failure to promptly identify and correct piping susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000, is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because if left uncorrected the finding could become a more significant event. The finding is of very low safety significance since the degradation of the system was identified and corrected prior to the safety function of the system being adversely impacted.
* The licensee failed to promptly implement timely corrective actions to operate and maintain the low pressure safety injection system as described in the Final Safety Analysis Report. Specifically, since 1997, the licensee utilized multiple analysis for evaluating degraded piping and pipe supports to evaluate acceptable void sizes. These analysis utilized allowable stresses that exceeded the design criteria allowable stresses described in the facilities Final Safety Analysis Report for the low pressure safety injection system. The failure to implement timely corrective actions to restore and maintain the low pressure safety injection system as described in the file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                      Page 3 of 7 Final Safety Analysis Report is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because the Mitigating Systems Objective to ensure the availability, reliability, and capability is potentially affected when the system is maintained outside of its design criteria as described in the Final Safety Analysis Report.
The finding is of very low safety significance since the analysis used to assess the degraded condition ensured the system could perform its safety function.
Inspection Report# : 2002005(pdf)
Significance:      Sep 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Evaluation Resulting in the Failure of the Shutdown Cooling System The licensee failed to adequately address the capability of the shutdown cooling system to perform its safety function after identifying a degraded condition. This resulted in the failure of two shutdown cooling suction isolation valves to open during attempts to line up the plant for shutdown cooling. The associated inadequate operability evaluation was determined to be a violation of Technical Specification 6.8.1(a) and Administrative Procedure LI-102, "Corrective Action Process," Revision 1. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of the shutdown cooling system during shutdown plant conditions. NRC Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process," was utilized to characterize the significance of the issue. During the loss of shutdown cooling on March 23, 2002, multiple systems or components were available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and high-pressure safety injection system. This event did not result in any loss of instrumentation needed for safe shutdown and cooldown of the plant. Based on multiple success paths available for ensuring decay heat removal capability and inventory makeup capability, this event was characterized as having very low safety significance (Section 1R15).
Inspection Report# : 2002003(pdf)
Significance:      Apr 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain proper foreign material exclusion controls The inspectors identified a violation of Technical Specification 6.8.1 for the failure to maintain proper foreign material exclusion controls in accordance with Procedure UNT 007-059 while working to correct a piping misalignment at Check Valve MS-402A, on the Emergency Feedwater AB main steam supply line. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0628. The safety significance of this violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for foreign material in the main steam supply line affecting the ability of the Emergency Feedwater AB turbine to operate as required. This issue was determined to be of very low safety significance because there was no foreign material found in the main steam piping system during the final system closure inspection.
Inspection Report# : 2002002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                      Page 4 of 7 Barrier Integrity Significance:      Jan 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples (one in barrier integrity) associated with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The barrier integrity example included:
The licensee failed to promptly identify and correct a degraded condition resulting in exceeding the rated thermal power limit from February 1995 to March 2002. This condition, identified by the licensee in March 2002, introduced non-conservative excore neutron detector calibration errors which affected the high linear power level, high logarithmic power level, high local power density, and low departure from nucleate boiling ratio, reactor protection trip functions.
The failure to promptly identify and correct the overpower condition was determined to be a violation of the facility operating License NPF-38 and 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because it affected four reactor trip functions in a non-conservative manner, thus, potentially impacting the barrier cornerstone integrity. The finding is of very low safety significance since it was determined that the accident analysis, Chapter 15 of the Final Safety Analysis Report, bounded the non-conservative trip functions. This finding is also of very low safety significance since actual fuel barrier integrity was never challenged during the overpower condition.
Inspection Report# : 2002005(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to establish an adequate maintenance procedure The licensee failed to establish an adequate maintenance procedure to ensure Switchgear Ventilation Damper SVS-102 remained in its safe position during maintenance and after the switchgear ventilation system was returned to an operable condition. Specifically, the damper was worked over a two day period without the damper being gagged in its safety minimum open position. The switchgear ventilation system was returned to an operable condition on September 19, 2002, without the associated actuator having been connected or a gag installed to maintain the damper in the minimal open position. The failure to gag the damper or restore the damper to an operable condition would have prevented the damper from being able to perform its safety function (minimum open position) on a safety injection actuation signal. The failure to provide adequate work instructions to repair Ventilation Damper SVS-102 is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the barrier integrity objective, to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events, was affected. A Phase 3 review was performed that considered the potential impact the switchgear ventilation system could have on the control envelope. The NRC risk analyst considered both radiological and toxic gas atmosphere. This finding is of very low safety significance, in part, based on a redundant damper being operable and the short duration the condition actually existed.
Inspection Report# : 2002004(pdf)
Significance:      Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Low Pressure Safety Injection System file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                        Page 5 of 7 The licensee failed to maintain design control of the low pressure safety injection system, Train A, in accordance with the design basis, as described in the Final Safety Analysis Report, when installing a modification to mitigate adverse voiding conditions that have affected the system. The failure to maintain design control of the system resulted in loss of a Seismic Class 1, ASME Section III, Safety Class 2, barrier during post accident conditions. The failure to maintain design control of the system is a violation of 10 CFR Part 50, Appendix B, Criterion III. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue screens more than minor because the Barrier Integrity Objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events was potentially affected. The finding is of very low safety significance since only degradation of the radiological barrier function provided for the auxiliary building was affected.
Inspection Report# : 2002005(pdf)
Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions The inspectors identified that the licensee failed to promptly identify and correct a condition adverse to quality, resulting in repetitive failures of solenoid-operated control valves to properly operate. The failure of these valves resulted in loss of the primary containment isolation function for the fire protection system piping penetrating containment. This was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of Valves FP-601A and -
601B to isolate on a containment isolation signal. This finding did not result in an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment. In accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, this issue was characterized as having very low safety significance.
Inspection Report# : 2002003(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Poor Radiological Work Planning During the review of the licensee's Refueling Outage 11 exposure estimates and exposure performance data, the inspectors identified that the Radiation Work Permit 2002-1600, "Health Physics Surveys and Postings," total person-rem exceeded budgeted person-rem by greater than 50 percent (5.7 rem verses 3.5 rem). From a review of the job-in-progress review, the inspectors noted that additional exposure was due, in part, to a higher source term than planned and increased radiation protection support for lower cavity and steam generator work that was not well communicated to the radiation protection staff. Additionally, the licensee did not reevaluate the dose estimate for Radiation Work Permit 2002-1600, when it was known that the actual effective dose rate was higher than planned. The failure to file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                      Page 6 of 7 reevaluate and adjust an as low as is reasonably achievable (ALARA) dose estimate was a performance deficiency. The finding was more than minor because it was associated with an Occupational Radiation Safety cornerstone attribute (ALARA Planning) and affected the associated cornerstone objective. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because actual job dose was more than 5 person-rem, it exceeded the planned intended dose by more than 50 percent, and the station's 3-year rolling average collective dose was less than 135 person-rem.
Inspection Report# : 2002003(pdf)
Significance:        Apr 12, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to survey and control a high radiation area The licensee failed to survey and control a high radiation area correctly. When items containing radioactive material were placed into a trash holding area on the foot elevation of containment on April 7, 2002, the licensee failed to perform a radiation survey, in violation of 10 CFR 20.1501(a), to evaluate the additional hazard and to adjust the placement of the rope barricade and warning signs (posting). Consequently, radiation dose rates at the rope barricade and posting exceeded the dose rates allowed by Technical Specification 6.12 by a factor of two, demonstrating that the rope barricade and posting did not encompass and control the entire high radiation area. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item is documented in the licensee's corrective action program as Condition Report 2002-00726. The finding had a credible impact on safety because workers could have unknowingly worked in the high radiation area outside the barricades and postings. The occurrence involved the potential for individual workers unplanned, unintended doses resulting from actions or conditions contrary to licensee procedures, Technical Specifications or NRC regulations, which could have been significantly greater if people had worked in the area. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had only very low safety significance because it is not an ALARA finding, an overexposure, a situation involving a substantial potential for overexposure, or an item compromising the ability to assess dose.
Inspection Report# : 2002002(pdf)
Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Dec 20, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee's process to identify, prioritize, evaluate, and correct problems was generally effective during calender years 2001 and 2002. The team reviewed 250 condition reports that were opened or closed during the period and found, in general, that station personnel effectively identified, characterized, and prioritized problems. Some issues involving file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Waterford 3                                                                        Page 7 of 7 the evaluation and correction of degraded conditions were identified by the team. Most of these issues were associated with longstanding degraded conditions that were identified and corrected by the licensee during this period and included the following: (1) an untimely identification of a void condition in the containment spray system existing between April and September 2002, (2) inadequate extent of condition reviews to identify main steam flow venturi degradation which existed since 1995 and the deleterious affect an oil coating which existed since 1997 would have on electrical components associated with the emergency diesel generator, (3) the inappropriate use of engineering analyses that allowed piping supports to exceed design basis allowable stresses during postulated accidents with voids in the low pressure safety injection system since 1997, (4) an inadequate verification of the design adequacy of a plant modification to vent low pressure safety injection system voids installed in June 2002, and (5) untimely corrective actions which resulted in a forced shutdown to repair weld cracks in the charging system in March 2000. Most of these issues had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2002005(pdf)
Significance:      Jun 29, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct the cause of voiding in the low pressure safety injection system and to take effective corrective action to preclude repetition of this condition The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct the cause of voiding in the low-pressure safety injection system and to take effective corrective action to preclude repetition of this condition, which had existed for several months prior to the March 23, through April 17,2002, refueling outage (RF11). The licensee failed to correct the root cause of this condition during the refueling outage. A total of seven unplanned entries into Technical Specification 3.5.2 Action (a) were made as a result of voids large enough to render Low-Pressure Safety Injection System Train A or B inoperable during the period April 18 - June 14, 2002. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensee's corrective action program as Condition Report 2002-0818. This violation was determined to have greater than minor safety significance because it had a credible impact on safety and did affect the operability, availability, reliability, and function of the low-pressure safety injection system. This issue was determined to be of very low safety significance because only one train of the low-pressure safety injection system was affected at any one time and the Technical Specification allowed outage time was not exceeded.
Inspection Report# : 2002002(pdf)
Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            07/22/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                        Page 1 of 7 Waterford 3 2Q/2003 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2003 Identified By: NRC Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified for the failure to maintain and operate main generator seal oil backup differential pressure regulating Valve SO-308 in accordance with vendor recommendations. This condition resulted in a turbine trip and subsequent reactor power cutback on February 14, 2003. This self-revealing finding is greater than minor because it resulted in a perturbation in plant stability resulting in a reactor power cutback, similar to example 4.b in Appendix E of Manual Chapter 0612. The finding is of very low safety significance because, although it caused a plant transient, it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3).
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow an operating procedure The licensee failed to follow Operating Procedure OP-002-003, "Component Cooling Water System," Revision 13, following maintenance activities on Essential Chiller A. The failure to follow procedure resulted in Component Cooling Water Valve CC-305A being mispositioned on November 22, 2002, affecting operability of both Component Cooling Water System Train A and Essential Chiller AB. The failure to follow an operating procedure is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the mitigating systems objective to ensure the availability and capability of the component cooling water and essential chill water systems were affected. The finding is of very low safety significance since the mispositioned valve did not result in loss of safety function for a single train for greater than the Technical Specification allowed outage time. The condition was promptly identified and corrected by the licensee approximately 1.5 hours after Valve CC-305A was mispositioned.
Inspection Report# : 2002004(pdf)
Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                        Page 2 of 7 Three examples associated (two in Mitigating Systems) with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The mitigating systems examples included:
* The licensee failed to promptly identify and correct a degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997. The team found that the licensee failed to evaluate the cause of the oil intrusion until 2001, took no corrective actions in 2001 or 2002 to prevent the oil intrusion when the source was identified, and failed to fully evaluate the detrimental effects that the oil intrusion could pose to the electrical and electronic components. The failure to promptly identify and correct the degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997 was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. 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 since the degraded condition did not result in a loss of the emergency diesel generator safety function.
* On April 18, 2002 when the low pressure safety injection Train B was found voided, the licensee failed to identify that the containment spray system Train B would also be voided from similar plant conditions. The containment spray voiding was identified by the licensee on September 17, 2002, when abnormal indications were noted by operators during a surveillance. Action was then taken by the licensee to correct the degraded condition. However, the licensee failed to identify the degraded condition during previous opportunities.
The failure to promptly identify and correct the voided condition affecting containment spray Train B was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected the voided condition could impact the reliability of the containment spray system to perform its safety function during accident conditions. The finding is of very low safety significance since the licensee could demonstrate through analysis that the actual degraded condition found would not have prevented the system from performing its safety function during accident conditions.
Inspection Report# : 2002005(pdf)
Significance:      Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Untimeliness Two examples of failures to implement timely corrective actions to resolve degraded conditions were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.c). Two examples included:
* The licensee failed to promptly identify and correct piping connections susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000. In 1997, the licensee experienced a crack of the charging system header due to fatigue stress cracking and determined additional piping connections were susceptible. The piping connection that failed in March 2000 was identified as being susceptible to fatigue stress cracking, however, no corrective actions had been taken. The failure to promptly identify and correct piping susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000, is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because if left uncorrected the finding could become a more significant event. The finding is of very low safety significance since the degradation of the system was identified and corrected prior to the safety function of the system being adversely impacted.
* The licensee failed to promptly implement timely corrective actions to operate and maintain the low pressure safety injection system as described in the Final Safety Analysis Report. Specifically, since 1997, the licensee utilized multiple analysis for evaluating degraded piping and pipe supports to evaluate acceptable void sizes. These analysis utilized allowable stresses that exceeded the design criteria allowable stresses described in the facilities Final Safety Analysis Report for the low pressure safety injection system. The failure to implement timely corrective actions to restore and maintain the low pressure safety injection system as described in the file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                      Page 3 of 7 Final Safety Analysis Report is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because the Mitigating Systems Objective to ensure the availability, reliability, and capability is potentially affected when the system is maintained outside of its design criteria as described in the Final Safety Analysis Report.
The finding is of very low safety significance since the analysis used to assess the degraded condition ensured the system could perform its safety function.
Inspection Report# : 2002005(pdf)
Significance:      Sep 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Evaluation Resulting in the Failure of the Shutdown Cooling System The licensee failed to adequately address the capability of the shutdown cooling system to perform its safety function after identifying a degraded condition. This resulted in the failure of two shutdown cooling suction isolation valves to open during attempts to line up the plant for shutdown cooling. The associated inadequate operability evaluation was determined to be a violation of Technical Specification 6.8.1(a) and Administrative Procedure LI-102, "Corrective Action Process," Revision 1. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of the shutdown cooling system during shutdown plant conditions. NRC Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process," was utilized to characterize the significance of the issue. During the loss of shutdown cooling on March 23, 2002, multiple systems or components were available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and high-pressure safety injection system. This event did not result in any loss of instrumentation needed for safe shutdown and cooldown of the plant. Based on multiple success paths available for ensuring decay heat removal capability and inventory makeup capability, this event was characterized as having very low safety significance (Section 1R15).
Inspection Report# : 2002003(pdf)
Barrier Integrity Significance:      Jan 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples (one in barrier integrity) associated with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The barrier integrity example included:
The licensee failed to promptly identify and correct a degraded condition resulting in exceeding the rated thermal power limit from February 1995 to March 2002. This condition, identified by the licensee in March 2002, introduced non-conservative excore neutron detector calibration errors which affected the high linear power level, high logarithmic power level, high local power density, and low departure from nucleate boiling ratio, reactor protection trip functions.
The failure to promptly identify and correct the overpower condition was determined to be a violation of the facility operating License NPF-38 and 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                        10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                      Page 4 of 7 noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because it affected four reactor trip functions in a non-conservative manner, thus, potentially impacting the barrier cornerstone integrity. The finding is of very low safety significance since it was determined that the accident analysis, Chapter 15 of the Final Safety Analysis Report, bounded the non-conservative trip functions. This finding is also of very low safety significance since actual fuel barrier integrity was never challenged during the overpower condition.
Inspection Report# : 2002005(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to establish an adequate maintenance procedure The licensee failed to establish an adequate maintenance procedure to ensure Switchgear Ventilation Damper SVS-102 remained in its safe position during maintenance and after the switchgear ventilation system was returned to an operable condition. Specifically, the damper was worked over a two day period without the damper being gagged in its safety minimum open position. The switchgear ventilation system was returned to an operable condition on September 19, 2002, without the associated actuator having been connected or a gag installed to maintain the damper in the minimal open position. The failure to gag the damper or restore the damper to an operable condition would have prevented the damper from being able to perform its safety function (minimum open position) on a safety injection actuation signal. The failure to provide adequate work instructions to repair Ventilation Damper SVS-102 is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the barrier integrity objective, to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events, was affected. A Phase 3 review was performed that considered the potential impact the switchgear ventilation system could have on the control envelope. The NRC risk analyst considered both radiological and toxic gas atmosphere. This finding is of very low safety significance, in part, based on a redundant damper being operable and the short duration the condition actually existed.
Inspection Report# : 2002004(pdf)
Significance:      Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Low Pressure Safety Injection System The licensee failed to maintain design control of the low pressure safety injection system, Train A, in accordance with the design basis, as described in the Final Safety Analysis Report, when installing a modification to mitigate adverse voiding conditions that have affected the system. The failure to maintain design control of the system resulted in loss of a Seismic Class 1, ASME Section III, Safety Class 2, barrier during post accident conditions. The failure to maintain design control of the system is a violation of 10 CFR Part 50, Appendix B, Criterion III. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue screens more than minor because the Barrier Integrity Objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events was potentially affected. The finding is of very low safety significance since only degradation of the radiological barrier function provided for the auxiliary building was affected.
Inspection Report# : 2002005(pdf)
Significance:      Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                        Page 5 of 7 The inspectors identified that the licensee failed to promptly identify and correct a condition adverse to quality, resulting in repetitive failures of solenoid-operated control valves to properly operate. The failure of these valves resulted in loss of the primary containment isolation function for the fire protection system piping penetrating containment. This was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue affected the reactor safety cornerstone objective in that this event challenged critical safety functions of Valves FP-601A and -
601B to isolate on a containment isolation signal. This finding did not result in an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment. In accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, this issue was characterized as having very low safety significance.
Inspection Report# : 2002003(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Apr 25, 2003 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.8.1.a required radiation work permit requirement The team identified a noncited violation of Technical Specification 6.8.1.a because the licensee failed to follow radiation work permit requirements. Specifically, on April 21, 2003, operations personnel entered into an unsurveyed radiologically restricted area in an overhead area in the reactor auxiliary building without first contacting radiation protection personnel prior to entry. This finding is greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation from radioactive material). The team processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Jul 19, 2002 Identified By: NRC Item Type: FIN Finding Poor Radiological Work Planning During the review of the licensee's Refueling Outage 11 exposure estimates and exposure performance data, the inspectors identified that the Radiation Work Permit 2002-1600, "Health Physics Surveys and Postings," total person-rem exceeded budgeted person-rem by greater than 50 percent (5.7 rem verses 3.5 rem). From a review of the job-in-progress review, the inspectors noted that additional exposure was due, in part, to a higher source term than planned file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                        Page 6 of 7 and increased radiation protection support for lower cavity and steam generator work that was not well communicated to the radiation protection staff. Additionally, the licensee did not reevaluate the dose estimate for Radiation Work Permit 2002-1600, when it was known that the actual effective dose rate was higher than planned. The failure to reevaluate and adjust an as low as is reasonably achievable (ALARA) dose estimate was a performance deficiency. The finding was more than minor because it was associated with an Occupational Radiation Safety cornerstone attribute (ALARA Planning) and affected the associated cornerstone objective. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because actual job dose was more than 5 person-rem, it exceeded the planned intended dose by more than 50 percent, and the station's 3-year rolling average collective dose was less than 135 person-rem.
Inspection Report# : 2002003(pdf)
Public Radiation Safety Significance:        Apr 25, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Violaton of 10 CFR 71.5 for failure to placard a transport vehicle containing hazardous material The team identified a self-revealing noncited violation of 10 CFR 71.5 because the licensee failed to placard a transport vehicle containing hazardous material. On June 5, 2002, the licensee was informed by letter from the recipient that Radioactive Material Shipment 02-3047 arrived at it's destination with no radioactive placards on the transport vehicle as required for radioactive material labeled as Radioactive Yellow III. This finding is greater than minor because it was associated with one of the Public Radiation Safety Cornerstone attributes (transportation program) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation materials released into the public domain). The team processed the violation through the Public Radiation Safety Significance Determination Process because the finding involved an occurrence in the licensee's radioactive material transportation program that is contrary to NRC and DOT regulations. The finding was a radioactive material control issue that involved transportation. However, it did not exceed radiation limits, involve a breach of package during transit, involve a Certificate of Compliance issue, involve a low level burial ground nonconformance, and involve a failure to make notifications or provide emergency information; therefore, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Physical Protection Miscellaneous Significance: N/A Dec 20, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee's process to identify, prioritize, evaluate, and correct problems was generally effective during calender years 2001 and 2002. The team reviewed 250 condition reports that were opened or closed during the period and found, file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Waterford 3                                                                      Page 7 of 7 in general, that station personnel effectively identified, characterized, and prioritized problems. Some issues involving the evaluation and correction of degraded conditions were identified by the team. Most of these issues were associated with longstanding degraded conditions that were identified and corrected by the licensee during this period and included the following: (1) an untimely identification of a void condition in the containment spray system existing between April and September 2002, (2) inadequate extent of condition reviews to identify main steam flow venturi degradation which existed since 1995 and the deleterious affect an oil coating which existed since 1997 would have on electrical components associated with the emergency diesel generator, (3) the inappropriate use of engineering analyses that allowed piping supports to exceed design basis allowable stresses during postulated accidents with voids in the low pressure safety injection system since 1997, (4) an inadequate verification of the design adequacy of a plant modification to vent low pressure safety injection system voids installed in June 2002, and (5) untimely corrective actions which resulted in a forced shutdown to repair weld cracks in the charging system in March 2000. Most of these issues had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2002005(pdf)
Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          10/08/2003
 
3Q/2003 Inspection Findings - Waterford 3                                                                        Page 1 of 8 Waterford 3 3Q/2003 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2003 Identified By: NRC Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified for the failure to maintain and operate main generator seal oil backup differential pressure regulating Valve SO-308 in accordance with vendor recommendations. This condition resulted in a turbine trip and subsequent reactor power cutback on February 14, 2003. This self-revealing finding is greater than minor because it resulted in a perturbation in plant stability resulting in a reactor power cutback, similar to example 4.b in Appendix E of Manual Chapter 0612. The finding is of very low safety significance because, although it caused a plant transient, it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3).
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Station Blackout Coping Analysis The inspectors identified a noncited violation of 10 CFR 50.63 for the failure to maintain a station blackout coping analysis that adequately encompassed plant conditions prescribed by the station blackout recovery emergency operating procedure. This resulted in the failure to evaluate for a reactor coolant system cooldown to a 400 F cold leg temperature, as prescribed by procedure, since the coping analysis assumed the reactor coolant system cold leg would be maintained at 545 F during station blackout conditions.
This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was determined to be of very low safety significance because the deficiency was confirmed not to result in loss of the capability to cope with a station blackout per Generic Letter 91-18 guidance.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                          Page 2 of 8 Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Starting Air System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to translate into specifications, procedures, and instructions design criteria for the diesel generator air start system. This resulted in the failure to maintain each diesel generator air receiver capable of starting the diesel engine five times. This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective due to the degradation of the design basis capability of the starting air system. The significance of the finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Overcurrent Relay A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" was identified for the failure to maintain design control of an overcurrent relay. This resulted in the failure to maintain normally open contact gap distances in accordance with vendor specifications. This design control deficiency was determined to be the most probable cause for loss of power to a safety related bus on July 24 and July 27, 2003. The finding is greater than minor because it affected the reactor safety mitigating system corner stone and if left uncorrected the finding could become a more significant safety concern. The significance of the finding was determined to be of very low safety significance because the deficiency did not result in the loss of safety-related equipment for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003006(pdf)
Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow an operating procedure The licensee failed to follow Operating Procedure OP-002-003, "Component Cooling Water System," Revision 13, following maintenance activities on Essential Chiller A. The failure to follow procedure resulted in Component Cooling Water Valve CC-305A being mispositioned on November 22, 2002, affecting operability of both Component Cooling Water System Train A and Essential Chiller AB. The failure to follow an operating procedure is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the mitigating systems objective to ensure the availability and capability of the component cooling water and essential chill water systems were affected. The finding is of very low safety significance since the mispositioned valve did not result in loss of safety function for a single train for greater than the Technical Specification allowed outage time. The condition was promptly identified and corrected by the licensee approximately 1.5 hours after Valve CC-305A was mispositioned.
Inspection Report# : 2002004(pdf)
Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                        Page 3 of 8 Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples associated (two in Mitigating Systems) with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The mitigating systems examples included:
* The licensee failed to promptly identify and correct a degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997. The team found that the licensee failed to evaluate the cause of the oil intrusion until 2001, took no corrective actions in 2001 or 2002 to prevent the oil intrusion when the source was identified, and failed to fully evaluate the detrimental effects that the oil intrusion could pose to the electrical and electronic components.
The failure to promptly identify and correct the degraded condition resulting in the electrical and electronic components inside Emergency Diesel Generator B control cabinet being subjected to oil intrusion since 1997 was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. 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 since the degraded condition did not result in a loss of the emergency diesel generator safety function.
* On April 18, 2002 when the low pressure safety injection Train B was found voided, the licensee failed to identify that the containment spray system Train B would also be voided from similar plant conditions. The containment spray voiding was identified by the licensee on September 17, 2002, when abnormal indications were noted by operators during a surveillance. Action was then taken by the licensee to correct the degraded condition. However, the licensee failed to identify the degraded condition during previous opportunities.
The failure to promptly identify and correct the voided condition affecting containment spray Train B was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected the voided condition could impact the reliability of the containment spray system to perform its safety function during accident conditions. The finding is of very low safety significance since the licensee could demonstrate through analysis that the actual degraded condition found would not have prevented the system from performing its safety function during accident conditions.
Inspection Report# : 2002005(pdf)
Significance:        Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Untimeliness Two examples of failures to implement timely corrective actions to resolve degraded conditions were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.c). Two examples included:
* The licensee failed to promptly identify and correct piping connections susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000. In 1997, the licensee experienced a crack of the charging system header due to fatigue stress cracking and determined additional piping connections were susceptible. The piping connection that failed in March 2000 was identified as being susceptible to fatigue stress cracking, however, no corrective actions had been taken.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                        Page 4 of 8 The failure to promptly identify and correct piping susceptible to fatigue stress cracking resulting in an unisolable leak from the charging system header on March 6, 2000, is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because if left uncorrected the finding could become a more significant event. The finding is of very low safety significance since the degradation of the system was identified and corrected prior to the safety function of the system being adversely impacted.
* The licensee failed to promptly implement timely corrective actions to operate and maintain the low pressure safety injection system as described in the Final Safety Analysis Report. Specifically, since 1997, the licensee utilized multiple analysis for evaluating degraded piping and pipe supports to evaluate acceptable void sizes. These analysis utilized allowable stresses that exceeded the design criteria allowable stresses described in the facilities Final Safety Analysis Report for the low pressure safety injection system.
The failure to implement timely corrective actions to restore and maintain the low pressure safety injection system as described in the Final Safety Analysis Report is a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The finding is greater than minor because the Mitigating Systems Objective to ensure the availability, reliability, and capability is potentially affected when the system is maintained outside of its design criteria as described in the Final Safety Analysis Report. The finding is of very low safety significance since the analysis used to assess the degraded condition ensured the system could perform its safety function.
Inspection Report# : 2002005(pdf)
Barrier Integrity Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Switchgear Ventilation System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to maintain design control of the switchgear ventilation system. This resulted in a potential common mode failure of safety related Dampers SVS-101 and SVS-102, due to loss of the nonsafety-related instrument air system.
The finding is greater than minor because if left uncorrected the finding could become a more significant safety concern. The significance of the finding, which is under the Barrier Integrity cornerstone, was determined to be of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003006(pdf)
Significance:        Jan 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples (one in barrier integrity) associated with failures to adequately evaluate the extent of conditions file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                      Page 5 of 8 adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The barrier integrity example included:
The licensee failed to promptly identify and correct a degraded condition resulting in exceeding the rated thermal power limit from February 1995 to March 2002. This condition, identified by the licensee in March 2002, introduced non-conservative excore neutron detector calibration errors which affected the high linear power level, high logarithmic power level, high local power density, and low departure from nucleate boiling ratio, reactor protection trip functions.
The failure to promptly identify and correct the overpower condition was determined to be a violation of the facility operating License NPF-38 and 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because it affected four reactor trip functions in a non-conservative manner, thus, potentially impacting the barrier cornerstone integrity. The finding is of very low safety significance since it was determined that the accident analysis, Chapter 15 of the Final Safety Analysis Report, bounded the non-conservative trip functions. This finding is also of very low safety significance since actual fuel barrier integrity was never challenged during the overpower condition.
Inspection Report# : 2002005(pdf)
Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to establish an adequate maintenance procedure The licensee failed to establish an adequate maintenance procedure to ensure Switchgear Ventilation Damper SVS-102 remained in its safe position during maintenance and after the switchgear ventilation system was returned to an operable condition. Specifically, the damper was worked over a two day period without the damper being gagged in its safety minimum open position. The switchgear ventilation system was returned to an operable condition on September 19, 2002, without the associated actuator having been connected or a gag installed to maintain the damper in the minimal open position. The failure to gag the damper or restore the damper to an operable condition would have prevented the damper from being able to perform its safety function (minimum open position) on a safety injection actuation signal. The failure to provide adequate work instructions to repair Ventilation Damper SVS-102 is a violation of Technical Specification 6.8.1(a). This finding is greater than minor because the barrier integrity objective, to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events, was affected. A Phase 3 review was performed that considered the potential impact the switchgear ventilation system could have on the control envelope. The NRC risk analyst considered both radiological and toxic gas atmosphere. This finding is of very low safety significance, in part, based on a redundant damper being operable and the short duration the condition actually existed.
Inspection Report# : 2002004(pdf)
Significance:      Dec 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Low Pressure Safety Injection System The licensee failed to maintain design control of the low pressure safety injection system, Train A, in accordance with the design basis, as described in the Final Safety Analysis Report, when installing a modification to mitigate adverse voiding conditions that have affected the system. The failure to maintain design control of the system resulted in loss of a Seismic Class 1, ASME Section III, Safety Class 2, barrier during post accident conditions.
The failure to maintain design control of the system is a violation of 10 CFR Part 50, Appendix B, Criterion III. This file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                        Page 6 of 8 violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This issue screens more than minor because the Barrier Integrity Objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events was potentially affected. The finding is of very low safety significance since only degradation of the radiological barrier function provided for the auxiliary building was affected.
Inspection Report# : 2002005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Apr 25, 2003 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.8.1.a required radiation work permit requirement The team identified a noncited violation of Technical Specification 6.8.1.a because the licensee failed to follow radiation work permit requirements. Specifically, on April 21, 2003, operations personnel entered into an unsurveyed radiologically restricted area in an overhead area in the reactor auxiliary building without first contacting radiation protection personnel prior to entry. This finding is greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation from radioactive material). The team processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Public Radiation Safety Significance:      Apr 25, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Violaton of 10 CFR 71.5 for failure to placard a transport vehicle containing hazardous material The team identified a self-revealing noncited violation of 10 CFR 71.5 because the licensee failed to placard a transport vehicle containing hazardous material. On June 5, 2002, the licensee was informed by letter from the recipient that Radioactive Material Shipment 02-3047 arrived at it's destination with no radioactive placards on the transport vehicle file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3                                                                      Page 7 of 8 as required for radioactive material labeled as Radioactive Yellow III. This finding is greater than minor because it was associated with one of the Public Radiation Safety Cornerstone attributes (transportation program) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation materials released into the public domain). The team processed the violation through the Public Radiation Safety Significance Determination Process because the finding involved an occurrence in the licensee's radioactive material transportation program that is contrary to NRC and DOT regulations. The finding was a radioactive material control issue that involved transportation. However, it did not exceed radiation limits, involve a breach of package during transit, involve a Certificate of Compliance issue, involve a low level burial ground nonconformance, and involve a failure to make notifications or provide emergency information; therefore, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Physical Protection Significance: N/A Jul 15, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Significance: N/A Dec 20, 2002 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The licensee's process to identify, prioritize, evaluate, and correct problems was generally effective during calender years 2001 and 2002. The team reviewed 250 condition reports that were opened or closed during the period and found, in general, that station personnel effectively identified, characterized, and prioritized problems. Some issues involving the evaluation and correction of degraded conditions were identified by the team. Most of these issues were associated with longstanding degraded conditions that were identified and corrected by the licensee during this period and included the following: (1) an untimely identification of a void condition in the containment spray system existing between April and September 2002, (2) inadequate extent of condition reviews to identify main steam flow venturi degradation which existed since 1995 and the deleterious affect an oil coating which existed since 1997 would have on electrical components associated with the emergency diesel generator, (3) the inappropriate use of engineering analyses that allowed piping supports to exceed design basis allowable stresses during postulated accidents with voids in the low pressure safety injection system since 1997, (4) an inadequate verification of the design adequacy of a plant modification to vent low pressure safety injection system voids installed in June 2002, and (5) untimely corrective actions which resulted in a forced shutdown to repair weld cracks in the charging system in March 2000. Most of these issues had cross-cutting aspects in the area of problem identification and resolution.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Waterford 3              Page 8 of 8 Inspection Report# : 2002005(pdf)
Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html 01/12/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                        Page 1 of 7 Waterford 3 4Q/2003 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2003 Identified By: NRC Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified for the failure to maintain and operate main generator seal oil backup differential pressure regulating Valve SO-308 in accordance with vendor recommendations. This condition resulted in a turbine trip and subsequent reactor power cutback on February 14, 2003. This self-revealing finding is greater than minor because it resulted in a perturbation in plant stability resulting in a reactor power cutback, similar to example 4.b in Appendix E of Manual Chapter 0612. The finding is of very low safety significance because, although it caused a plant transient, it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3).
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of Voiding Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to establish adequate corrective actions to prevent recurrence of voiding conditions affecting the operability of the low pressure safety injection system following shutdown cooling operations. This finding is greater than minor because it affected the mitigating system objective to ensure the reliability and availability of the low pressure safety injection system to respond to an initiating event. The problem if left uncorrected would become a more significant safety concern. The significance of this finding was determined to be of very low safety significance because low pressure safety injection Train B was inoperable for less than the Technical Specification allowed outage time and Train A was determined to be degraded but operable in accordance with Generic Letter 91-18 guidance.
Inspection Report# : 2003007(pdf)
Significance: TBD Dec 31, 2003 Identified By: Self Disclosing Item Type: AV Apparent Violation Failure to establish appropriate instructions and implement those instructions A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                          Page 2 of 7 Drawings," was identified for the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel line for Train A emergency diesel generator in May 2003. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test. This finding is unresolved pending completion of a significance determination. The finding was greater than minor because it directly impacted the availability and reliability of an emergency diesel generator which is used to mitigate the loss of AC power to the respective safety related bus. The finding was determined to have a potential safety significance greater than very low significance because the failure resulted in an actual loss of the safety function of the Train A emergency diesel generator for an extended period of time.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Station Blackout Coping Analysis The inspectors identified a noncited violation of 10 CFR 50.63 for the failure to maintain a station blackout coping analysis that adequately encompassed plant conditions prescribed by the station blackout recovery emergency operating procedure. This resulted in the failure to evaluate for a reactor coolant system cooldown to a 400 F cold leg temperature, as prescribed by procedure, since the coping analysis assumed the reactor coolant system cold leg would be maintained at 545 F during station blackout conditions.
This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was determined to be of very low safety significance because the deficiency was confirmed not to result in loss of the capability to cope with a station blackout per Generic Letter 91-18 guidance.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Starting Air System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to translate into specifications, procedures, and instructions design criteria for the diesel generator air start system. This resulted in the failure to maintain each diesel generator air receiver capable of starting the diesel engine five times. This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective due to the degradation of the design basis capability of the starting air system. The significance of the finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Overcurrent Relay file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                        Page 3 of 7 A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" was identified for the failure to maintain design control of an overcurrent relay. This resulted in the failure to maintain normally open contact gap distances in accordance with vendor specifications. This design control deficiency was determined to be the most probable cause for loss of power to a safety related bus on July 24 and July 27, 2003. The finding is greater than minor because it affected the reactor safety mitigating system corner stone and if left uncorrected the finding could become a more significant safety concern. The significance of the finding was determined to be of very low safety significance because the deficiency did not result in the loss of safety-related equipment for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003006(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to test certain emergency diesel generator "B" mini-sequencer contacts.
The identified a violation of TS 6.8.1.f for failure to establish a procedure that implements a procedure to functionally test certain electrical circuits on the EDG mini-sequencer, which is relied upon for achieving shutdown in the event of a fire requiring control room evacuation and remote shutdown. Upon failure of this portion of the sequencer, automatic sequencing of certain components required for safe shutdown would be lost.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency lighting for supporting operator actions.
The team identified that the licensee had not provided sufficient emergency lighting for a safe shutdown of the plant follwoing a fire and evacuation of the control room.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions for deficiencies associated with the emergency lighting system.
Two examples:
: 1. The licensee failed to complete actions to correct aconditions adverse to fire protection, in that, they inappropriately cancelled a full-field verification test of their emergency lightting system.
: 2. The licensee failed to correct a deficiency in their methodology for determining if the emergency lighting system met the 10 CFR 50.65, Section (a)(1), maintenance rule goals.
Inspection Report# : 2003011(pdf)
Barrier Integrity file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                        Page 4 of 7 Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Controls of MSIVs The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing main steam isolation Valves 1 and 2. This performance deficiency contributed to both valves being declared inoperable due to system leaks creating a low pressure condition in the valve actuating systems. This finding is more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance of the functionality of containment. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2003007(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of PWSCC of Alloy 600 Material The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions resulting in recurrences of pressure boundary leakage due to primary water stress corrosion cracking of Alloy 600 reactor coolant system nozzles. This finding was greater than minor because it affected the reactor safety barrier integrity cornerstone objective for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Manual Chapter 0609 Significance determination process Phase 1 Screening Worksheet this performance deficiency affected the reactor coolant system barrier function requiring a Phase 2 analysis. The results of the Phase 2 and 3 analysis determined that this finding was of very low safety significance based on the cracks being axial in nature (does not contribute substantially to a loss of coolant accident) and the leaks resulted in a build up of only minor boric acid residue indicative of only trace amounts of through wall leakage. The leak rates identified were well within the capacity of a single charging pump.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Switchgear Ventilation System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to maintain design control of the switchgear ventilation system. This resulted in a potential common mode failure of safety related Dampers SVS-101 and SVS-102, due to loss of the nonsafety-related instrument air system.
The finding is greater than minor because if left uncorrected the finding could become a more significant safety concern. The significance of the finding, which is under the Barrier Integrity cornerstone, was determined to be of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                      Page 5 of 7 Significance:      Jan 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions Resulting from Inadequate Evaluations of Extent of Condition Three examples (one in barrier integrity) associated with failures to adequately evaluate the extent of conditions adverse to quality were identified. The failure to promptly identify and correct these degraded conditions was a violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section 4OA2.b). The barrier integrity example included:
The licensee failed to promptly identify and correct a degraded condition resulting in exceeding the rated thermal power limit from February 1995 to March 2002. This condition, identified by the licensee in March 2002, introduced non-conservative excore neutron detector calibration errors which affected the high linear power level, high logarithmic power level, high local power density, and low departure from nucleate boiling ratio, reactor protection trip functions.
The failure to promptly identify and correct the overpower condition was determined to be a violation of the facility operating License NPF-38 and 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because it affected four reactor trip functions in a non-conservative manner, thus, potentially impacting the barrier cornerstone integrity. The finding is of very low safety significance since it was determined that the accident analysis, Chapter 15 of the Final Safety Analysis Report, bounded the non-conservative trip functions. This finding is also of very low safety significance since actual fuel barrier integrity was never challenged during the overpower condition.
Inspection Report# : 2002005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Oct 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Barricade a High Radiation Area The inspector identified a noncited violation of Technical Specification 6.12.1 because Entergy failed to barricade a high radiation area. Specifically, on October 27, 2003, the inspector observed that the high radiation area rope barricading the regenitive heat exchanger room was stretched across the entrance way at a height of approximately 79 inches, which would not obstruct the entry of station workers. General area radiation levels within the room were as high as 420 millirem per hour. The finding is in Entergy's corrective action program as Condition Report CR-WF3-2003-03164. The finding is greater than minor because it affected the Occupational Radiation Safety cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation and the finding is associated with the cornerstone attribute (Program & Process). The finding involved an individual's potential for unplanned or unintended dose. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with ALARA planning or work controls, there was no overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                        Page 6 of 7 Inspection Report# : 2003007(pdf)
Significance:      Apr 25, 2003 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.8.1.a required radiation work permit requirement The team identified a noncited violation of Technical Specification 6.8.1.a because the licensee failed to follow radiation work permit requirements. Specifically, on April 21, 2003, operations personnel entered into an unsurveyed radiologically restricted area in an overhead area in the reactor auxiliary building without first contacting radiation protection personnel prior to entry. This finding is greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation from radioactive material). The team processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Public Radiation Safety Significance:      Apr 25, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Violaton of 10 CFR 71.5 for failure to placard a transport vehicle containing hazardous material The team identified a self-revealing noncited violation of 10 CFR 71.5 because the licensee failed to placard a transport vehicle containing hazardous material. On June 5, 2002, the licensee was informed by letter from the recipient that Radioactive Material Shipment 02-3047 arrived at it's destination with no radioactive placards on the transport vehicle as required for radioactive material labeled as Radioactive Yellow III. This finding is greater than minor because it was associated with one of the Public Radiation Safety Cornerstone attributes (transportation program) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation materials released into the public domain). The team processed the violation through the Public Radiation Safety Significance Determination Process because the finding involved an occurrence in the licensee's radioactive material transportation program that is contrary to NRC and DOT regulations. The finding was a radioactive material control issue that involved transportation. However, it did not exceed radiation limits, involve a breach of package during transit, involve a Certificate of Compliance issue, involve a low level burial ground nonconformance, and involve a failure to make notifications or provide emergency information; therefore, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Waterford 3                                                                  Page 7 of 7 Physical Protection Significance: N/A Jul 15, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\WAT3\wat3_pim.html                                                    04/22/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                                  Page 1 of 6 Waterford 3 1Q/2004 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Fuel Oil Storage Requirements The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the emergency diesel generating (EDG) system fuel oil storage requirements. This failure affected the ability of each emergency diesel generator to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offsite power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance because the licensee maintains additional diesel fuel oil onsite and replenishing the fuel oil via tanker truck, train, or barge is readily available based on the site being located in a heavy industrial corridor of Louisiana where there are many oil refineries and oil storage facilities.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct EDG Loading and Fuel Oil Consumption Analysis Deifciencies The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee inappropriately closed a corrective action requiring revisions to the EDG Loading and Fuel Oil Consumption analysis. The failure to adequately complete this corrective action resulted in the failure to maintain design control of the EDG fuel oil storage inventory requirements to ensure a 7-day post accident fuel oil inventory. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance because the licensee maintains additional diesel fuel oil onsite and replenishing the fuel oil via tanker truck, train, or barge is readily available based on the site being located in a heavy industrial corridor of Louisiana where there are many oil refineries and oil storage facilities.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of Voiding Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to establish adequate corrective actions to prevent recurrence of voiding conditions affecting the operability of the low pressure safety injection system following shutdown cooling operations. This finding is greater than minor because it affected the mitigating system objective to ensure the reliability and availability of the low pressure safety injection system to respond to an initiating event. The problem if left uncorrected would become a more significant safety concern. The significance of this finding was determined to be of very low safety significance because low pressure safety injection Train B was inoperable for less than the Technical Specification allowed outage time and Train A was determined to be degraded but operable in accordance with Generic Letter 91-18 guidance.
Inspection Report# : 2003007(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                                Page 2 of 6 Significance:        Dec 31, 2003 Identified By: NRC Item Type: VIO Violation Failure to establish appropriate instructions and implement those instructions Contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings, during the overhaul of Train A emergency diesel generator in May 2003, the licensee failed to establish adequate instructions to ensure proper installation of th fuel supply line of Train A emergency diesel generator. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Station Blackout Coping Analysis The inspectors identified a noncited violation of 10 CFR 50.63 for the failure to maintain a station blackout coping analysis that adequately encompassed plant conditions prescribed by the station blackout recovery emergency operating procedure. This resulted in the failure to evaluate for a reactor coolant system cooldown to a 400 F cold leg temperature, as prescribed by procedure, since the coping analysis assumed the reactor coolant system cold leg would be maintained at 545 F during station blackout conditions.
This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was determined to be of very low safety significance because the deficiency was confirmed not to result in loss of the capability to cope with a station blackout per Generic Letter 91-18 guidance.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Starting Air System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to translate into specifications, procedures, and instructions design criteria for the diesel generator air start system. This resulted in the failure to maintain each diesel generator air receiver capable of starting the diesel engine five times. This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective due to the degradation of the design basis capability of the starting air system. The significance of the finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Overcurrent Relay A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" was identified for the failure to maintain design control of an overcurrent relay. This resulted in the failure to maintain normally open contact gap distances in accordance with vendor specifications. This design control deficiency was determined to be the most probable cause for loss of power to a safety related bus on July 24 and July 27, 2003. The finding is greater than minor because it affected the reactor safety mitigating system corner stone and if left uncorrected the finding could become a more significant safety concern. The significance of the finding was determined to be of very low safety significance because the deficiency did not result in the loss of safety-related equipment for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003006(pdf)
Significance:        Aug 19, 2003 Identified By: NRC 07/14/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                              Page 3 of 6 Item Type: NCV NonCited Violation Failure to test certain emergency diesel generator "B" mini-sequencer contacts.
The identified a violation of TS 6.8.1.f for failure to establish a procedure that implements a procedure to functionally test certain electrical circuits on the EDG mini-sequencer, which is relied upon for achieving shutdown in the event of a fire requiring control room evacuation and remote shutdown. Upon failure of this portion of the sequencer, automatic sequencing of certain components required for safe shutdown would be lost.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency lighting for supporting operator actions.
The team identified that the licensee had not provided sufficient emergency lighting for a safe shutdown of the plant follwoing a fire and evacuation of the control room.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions for deficiencies associated with the emergency lighting system.
Two examples:
: 1. The licensee failed to complete actions to correct aconditions adverse to fire protection, in that, they inappropriately cancelled a full-field verification test of their emergency lightting system.
: 2. The licensee failed to correct a deficiency in their methodology for determining if the emergency lighting system met the 10 CFR 50.65, Section (a)(1), maintenance rule goals.
Inspection Report# : 2003011(pdf)
Barrier Integrity Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for the failure to promptly identify and correct a condition adverse to quality. Specifically, Entergy failed to replace known age-degraded O-rings affecting the main feedwater isolation valves in the year 2000 resulting in O-ring failure and inoperability of the Train A feedwater isolation valve on December 27, 2003.
The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Maintenance Instruction Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to establish appropriate instructions for corrective maintenance activities on Train A main feedwater isolation valve on December 27, 2003. This resulted in the failure to establish appropriate torque specifications to ensure adequate O-ring compression that ultimately led to an O-ring failure and inoperability of the isolation valve on January 3, 2004. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of 07/14/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                            Page 4 of 6 the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Controls of MSIVs The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing main steam isolation Valves 1 and 2. This performance deficiency contributed to both valves being declared inoperable due to system leaks creating a low pressure condition in the valve actuating systems. This finding is more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance of the functionality of containment. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2003007(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of PWSCC of Alloy 600 Material The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions resulting in recurrences of pressure boundary leakage due to primary water stress corrosion cracking of Alloy 600 reactor coolant system nozzles. This finding was greater than minor because it affected the reactor safety barrier integrity cornerstone objective for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Manual Chapter 0609 Significance determination process Phase 1 Screening Worksheet this performance deficiency affected the reactor coolant system barrier function requiring a Phase 2 analysis. The results of the Phase 2 and 3 analysis determined that this finding was of very low safety significance based on the cracks being axial in nature (does not contribute substantially to a loss of coolant accident) and the leaks resulted in a build up of only minor boric acid residue indicative of only trace amounts of through wall leakage. The leak rates identified were well within the capacity of a single charging pump.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Switchgear Ventilation System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to maintain design control of the switchgear ventilation system. This resulted in a potential common mode failure of safety related Dampers SVS-101 and SVS-102, due to loss of the nonsafety-related instrument air system. The finding is greater than minor because if left uncorrected the finding could become a more significant safety concern. The significance of the finding, which is under the Barrier Integrity cornerstone, was determined to be of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003006(pdf)
Emergency Preparedness Occupational Radiation Safety 07/14/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                                Page 5 of 6 Significance:        Oct 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Barricade a High Radiation Area The inspector identified a noncited violation of Technical Specification 6.12.1 because Entergy failed to barricade a high radiation area.
Specifically, on October 27, 2003, the inspector observed that the high radiation area rope barricading the regenitive heat exchanger room was stretched across the entrance way at a height of approximately 79 inches, which would not obstruct the entry of station workers. General area radiation levels within the room were as high as 420 millirem per hour. The finding is in Entergy's corrective action program as Condition Report CR-WF3-2003-03164. The finding is greater than minor because it affected the Occupational Radiation Safety cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation and the finding is associated with the cornerstone attribute (Program & Process). The finding involved an individual's potential for unplanned or unintended dose. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with ALARA planning or work controls, there was no overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003007(pdf)
Significance:        Apr 25, 2003 Identified By: NRC Item Type: NCV NonCited Violation Violation of Technical Specification 6.8.1.a required radiation work permit requirement The team identified a noncited violation of Technical Specification 6.8.1.a because the licensee failed to follow radiation work permit requirements. Specifically, on April 21, 2003, operations personnel entered into an unsurveyed radiologically restricted area in an overhead area in the reactor auxiliary building without first contacting radiation protection personnel prior to entry. This finding is greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation from radioactive material). The team processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf)
Public Radiation Safety Significance:        Apr 25, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Violaton of 10 CFR 71.5 for failure to placard a transport vehicle containing hazardous material The team identified a self-revealing noncited violation of 10 CFR 71.5 because the licensee failed to placard a transport vehicle containing hazardous material. On June 5, 2002, the licensee was informed by letter from the recipient that Radioactive Material Shipment 02-3047 arrived at it's destination with no radioactive placards on the transport vehicle as required for radioactive material labeled as Radioactive Yellow III. This finding is greater than minor because it was associated with one of the Public Radiation Safety Cornerstone attributes (transportation program) and the finding affected the associated cornerstone objective (to ensure the adequate protection of public health and safety from exposure to radiation materials released into the public domain). The team processed the violation through the Public Radiation Safety Significance Determination Process because the finding involved an occurrence in the licensee's radioactive material transportation program that is contrary to NRC and DOT regulations. The finding was a radioactive material control issue that involved transportation.
However, it did not exceed radiation limits, involve a breach of package during transit, involve a Certificate of Compliance issue, involve a low level burial ground nonconformance, and involve a failure to make notifications or provide emergency information; therefore, the violation had no more than very low safety significance.
Inspection Report# : 2003008(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Waterford 3                                                                                          Page 6 of 6 Physical Protection Significance: N/A Jul 15, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Last modified : May 05, 2004 07/14/2004
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                          Page 1 of 6 Waterford 3 2Q/2004 Plant Inspection Findings Initiating Events Significance:        Jun 26, 2004 Identified By: NRC Item Type: FIN Finding Improper Maintenance Activities resulting in Plant Down Power A self-revealing finding was identified involving improper installation of an O-ring for Emergency Header Check Valve EH-1285. This resulted in an unisolable hydraulic fluid leak in the main turbine electro-hydraulic control system. Entergy elected to reduce reactor power to less than 20 percent and manually trip the main turbine on February 14, 2004. This self-revealing finding is greater than minor because it is associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. The human performance attribute was affected in that the performance deficiency resulted in a perturbation in plant stability by reducing reactor power to less than 20 percent. Although the unisolable hydraulic leak resulted in a plant transient, the finding is of very low safety significance because it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004003(pdf)
Mitigating Systems Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct instances where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions. In one case, the licensee failed to properly address system operability and, for a two-week period, actual valve operability was unknown. The issue was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was determined to be of very low risk significance because each issue: was not a design or qualification deficiency; did not result in the loss of a safety system; did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours; and was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensee's corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Fuel Oil Storage Requirements The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the emergency diesel generating (EDG) system fuel oil storage requirements. This failure affected the ability of each emergency diesel generator to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offsite power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance because the licensee maintains additional diesel fuel oil onsite and replenishing the fuel oil via tanker truck, train, or barge is readily available based on the site being located in a heavy industrial corridor of Louisiana where there are many oil refineries and oil storage facilities.
Inspection Report# : 2004002(pdf)
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                          Page 2 of 6 Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct EDG Loading and Fuel Oil Consumption Analysis Deifciencies The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee inappropriately closed a corrective action requiring revisions to the EDG Loading and Fuel Oil Consumption analysis. The failure to adequately complete this corrective action resulted in the failure to maintain design control of the EDG fuel oil storage inventory requirements to ensure a 7-day post accident fuel oil inventory. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance because the licensee maintains additional diesel fuel oil onsite and replenishing the fuel oil via tanker truck, train, or barge is readily available based on the site being located in a heavy industrial corridor of Louisiana where there are many oil refineries and oil storage facilities.
Inspection Report# : 2004002(pdf)
Significance:        Jan 05, 2004 Identified By: NRC Item Type: VIO Violation Failure to establish appropriate instructions and implement those instructions Contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings, during the overhaul of Train A emergency diesel generator in May 2003, the licensee failed to establish adequate instructions to ensure proper installation of th fuel supply line of Train A emergency diesel generator. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test.
Inspection Report# : 2003007(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of Voiding Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to establish adequate corrective actions to prevent recurrence of voiding conditions affecting the operability of the low pressure safety injection system following shutdown cooling operations.
This finding is greater than minor because it affected the mitigating system objective to ensure the reliability and availability of the low pressure safety injection system to respond to an initiating event. The problem if left uncorrected would become a more significant safety concern. The significance of this finding was determined to be of very low safety significance because low pressure safety injection Train B was inoperable for less than the Technical Specification allowed outage time and Train A was determined to be degraded but operable in accordance with Generic Letter 91-18 guidance.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Station Blackout Coping Analysis The inspectors identified a noncited violation of 10 CFR 50.63 for the failure to maintain a station blackout coping analysis that adequately encompassed plant conditions prescribed by the station blackout recovery emergency operating procedure. This resulted in the failure to evaluate for a reactor coolant system cooldown to a 400 F cold leg temperature, as prescribed by procedure, since the coping analysis assumed the reactor coolant system cold leg would be maintained at 545 F during station blackout conditions.
This finding is greater than minor because it affected the reactor safety mitigating system cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was determined to be of very low safety significance because the deficiency was confirmed not to result in loss of the capability to cope with a station blackout per Generic Letter 91-18 guidance.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Starting Air System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to translate into specifications, procedures, and instructions design criteria for the diesel generator air start system. This resulted in the failure to maintain each diesel generator air receiver capable of starting the diesel engine five times. This finding is greater than minor because it affected the reactor safety mitigating
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                        Page 3 of 6 system cornerstone objective due to the degradation of the design basis capability of the starting air system. The significance of the finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2003006(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Overcurrent Relay A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" was identified for the failure to maintain design control of an overcurrent relay. This resulted in the failure to maintain normally open contact gap distances in accordance with vendor specifications.
This design control deficiency was determined to be the most probable cause for loss of power to a safety related bus on July 24 and July 27, 2003. The finding is greater than minor because it affected the reactor safety mitigating system corner stone and if left uncorrected the finding could become a more significant safety concern. The significance of the finding was determined to be of very low safety significance because the deficiency did not result in the loss of safety-related equipment for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003006(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to test certain emergency diesel generator "B" mini-sequencer contacts.
The identified a violation of TS 6.8.1.f for failure to establish a procedure that implements a procedure to functionally test certain electrical circuits on the EDG mini-sequencer, which is relied upon for achieving shutdown in the event of a fire requiring control room evacuation and remote shutdown.
Upon failure of this portion of the sequencer, automatic sequencing of certain components required for safe shutdown would be lost.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate emergency lighting for supporting operator actions.
The team identified that the licensee had not provided sufficient emergency lighting for a safe shutdown of the plant follwoing a fire and evacuation of the control room.
Inspection Report# : 2003011(pdf)
Significance:        Aug 19, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions for deficiencies associated with the emergency lighting system.
Two examples:
: 1. The licensee failed to complete actions to correct aconditions adverse to fire protection, in that, they inappropriately cancelled a full-field verification test of their emergency lightting system.
: 2. The licensee failed to correct a deficiency in their methodology for determining if the emergency lighting system met the 10 CFR 50.65, Section (a)
(1), maintenance rule goals.
Inspection Report# : 2003011(pdf)
Barrier Integrity Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                      Page 4 of 6 to calculate the control room operator dose for design basis accident conditions involving radiological releases. This failure resulted in significantly underestimating the actual dose to operators. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope and was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as condition report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for the failure to promptly identify and correct a condition adverse to quality. Specifically, Entergy failed to replace known age-degraded O-rings affecting the main feedwater isolation valves in the year 2000 resulting in O-ring failure and inoperability of the Train A feedwater isolation valve on December 27, 2003. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Maintenance Instruction Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to establish appropriate instructions for corrective maintenance activities on Train A main feedwater isolation valve on December 27, 2003. This resulted in the failure to establish appropriate torque specifications to ensure adequate O-ring compression that ultimately led to an O-ring failure and inoperability of the isolation valve on January 3, 2004. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Controls of MSIVs The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing main steam isolation Valves 1 and 2. This performance deficiency contributed to both valves being declared inoperable due to system leaks creating a low pressure condition in the valve actuating systems. This finding is more than minor because it affected the Barrier Integrity
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                      Page 5 of 6 Cornerstone objective of providing reasonable assurance of the functionality of containment. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2003007(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of PWSCC of Alloy 600 Material The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions resulting in recurrences of pressure boundary leakage due to primary water stress corrosion cracking of Alloy 600 reactor coolant system nozzles. This finding was greater than minor because it affected the reactor safety barrier integrity cornerstone objective for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Manual Chapter 0609 Significance determination process Phase 1 Screening Worksheet this performance deficiency affected the reactor coolant system barrier function requiring a Phase 2 analysis. The results of the Phase 2 and 3 analysis determined that this finding was of very low safety significance based on the cracks being axial in nature (does not contribute substantially to a loss of coolant accident) and the leaks resulted in a build up of only minor boric acid residue indicative of only trace amounts of through wall leakage. The leak rates identified were well within the capacity of a single charging pump.
Inspection Report# : 2003007(pdf)
Significance:        Sep 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Switchgear Ventilation System The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to maintain design control of the switchgear ventilation system. This resulted in a potential common mode failure of safety related Dampers SVS-101 and SVS-102, due to loss of the nonsafety-related instrument air system. The finding is greater than minor because if left uncorrected the finding could become a more significant safety concern. The significance of the finding, which is under the Barrier Integrity cornerstone, was determined to be of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003006(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Oct 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Barricade a High Radiation Area The inspector identified a noncited violation of Technical Specification 6.12.1 because Entergy failed to barricade a high radiation area. Specifically, on October 27, 2003, the inspector observed that the high radiation area rope barricading the regenitive heat exchanger room was stretched across the entrance way at a height of approximately 79 inches, which would not obstruct the entry of station workers. General area radiation levels within the room were as high as 420 millirem per hour. The finding is in Entergy's corrective action program as Condition Report CR-WF3-2003-03164. The finding is greater than minor because it affected the Occupational Radiation Safety cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation and the finding is associated with the cornerstone attribute (Program & Process). The finding involved an individual's potential for unplanned or unintended dose. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with ALARA planning or work controls, there was no overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003007(pdf)
Public Radiation Safety
 
2Q/2004 Inspection Findings - Waterford 3                                                                                                      Page 6 of 6 Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A May 21, 2004 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the licensee's process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2004006(pdf)
Last modified : September 08, 2004
 
3Q/2004 Inspection Findings - Waterford 3                                                                                              Page 1 of 6 Waterford 3 3Q/2004 Plant Inspection Findings Initiating Events Significance:        Jun 26, 2004 Identified By: NRC Item Type: FIN Finding Improper Maintenance Activities resulting in Plant Down Power A self-revealing finding was identified involving improper installation of an O-ring for Emergency Header Check Valve EH-1285. This resulted in an unisolable hydraulic fluid leak in the main turbine electro-hydraulic control system. Entergy elected to reduce reactor power to less than 20 percent and manually trip the main turbine on February 14, 2004. This self-revealing finding is greater than minor because it is associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. The human performance attribute was affected in that the performance deficiency resulted in a perturbation in plant stability by reducing reactor power to less than 20 percent. Although the unisolable hydraulic leak resulted in a plant transient, the finding is of very low safety significance because it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004003(pdf)
Mitigating Systems Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Environmental Qualification Records The inspectors identified a noncited violation of 10 CFR 50.49(j) for the failure to maintain an auditable record demonstrating that electric equipment important to safety is environmentally qualified for its intended application. Specifically, it was identified that nonconservative temperature profiles were utilized to calculate the qualified life of ASCO NP8300 series solenoid-operated valves. The finding was more than minor since if left uncorrected it would become a more significant safety concern. Specifically, the failure to maintain electrical equipment in an environmentally qualified configuration could adversely impact the ability of such mitigating equipment to perform its safety function during design-basis accident conditions. This finding was of very low safety significance since additional analysis demonstrated that affected electrical equipment currently installed in the plant was environmentally qualifiable. Therefore, this deficiency did not result in any loss of affected equipment safety function.
Inspection Report# : 2004004(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Safety Injection Sump Recirculation Piping The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the containment safety injection sump recirculation piping. This deficiency resulted in inappropriately maintaining a section of the piping void of water, potentially affecting the operability of the high-pressure safety injection and containment spray pumps during postulated design-basis accident conditions following a recirculation actuation signal. This finding was more than minor because it potentially affected the mitigating system cornerstone objective of ensuring the capability of the high-pressure safety injection and containment spray systems to perform their design-basis functions. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004004(pdf)
Significance:        May 21, 2004 Identified By: NRC
 
3Q/2004 Inspection Findings - Waterford 3                                                                                                Page 2 of 6 Item Type: NCV NonCited Violation Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct instances where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions. In one case, the licensee failed to properly address system operability and, for a two-week period, actual valve operability was unknown. The issue was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was determined to be of very low risk significance because each issue: was not a design or qualification deficiency; did not result in the loss of a safety system; did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours; and was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensee's corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Fuel Oil Storage Requirements The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the emergency diesel generating (EDG) system fuel oil storage requirements. This failure affected the ability of each emergency diesel generator to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offsite power and a design-bases accident.
This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency ac power to respond to initiating events to prevent undesirable consequences. This finding was evaluated using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the mitigating systems cornerstone. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in a loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct EDG Loading and Fuel Oil Consumption Analysis Deifciencies The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee inappropriately closed a corrective action requiring revisions to the EDG Loading and Fuel Oil Consumption analysis. The failure to adequately complete this corrective action resulted in the failure to maintain design control of the EDG fuel oil storage inventory requirements to ensure a 7-day post accident fuel oil inventory. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency ac power to respond to initiating events to prevent undesirable consequences. This finding was evaluated using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the mitigating systems cornerstone. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in a loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004002(pdf)
Significance:        Jan 05, 2004 Identified By: NRC Item Type: VIO Violation Failure to establish appropriate instructions and implement those instructions Contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings, during the overhaul of Train A emergency diesel generator in May 2003, the licensee failed to establish adequate instructions to ensure proper installation of th fuel supply line of Train A emergency diesel generator. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test.
Inspection Report# : 2003007(pdf)
Inspection Report# : 2004008(pdf)
 
3Q/2004 Inspection Findings - Waterford 3                                                                                            Page 3 of 6 Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of Voiding Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to establish adequate corrective actions to prevent recurrence of voiding conditions affecting the operability of the low pressure safety injection system following shutdown cooling operations. This finding is greater than minor because it affected the mitigating system objective to ensure the reliability and availability of the low pressure safety injection system to respond to an initiating event. The problem if left uncorrected would become a more significant safety concern. The significance of this finding was determined to be of very low safety significance because low pressure safety injection Train B was inoperable for less than the Technical Specification allowed outage time and Train A was determined to be degraded but operable in accordance with Generic Letter 91-18 guidance.
Inspection Report# : 2003007(pdf)
Barrier Integrity Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Pevent Recurrence of Main Steam Isolation Valve Failures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to determine the cause and preclude recurrence of main steam isolation solenoid-operated dump valve failures. This failure affected the primary containment isolation function for the main steam system isolation valves. The primary cause of this finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because if left uncorrected the finding could become a more safety significant concern. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time. The valve was repaired and returned to service.
Inspection Report# : 2004004(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions Affecting the Emergency Feedwater System A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified when the valve failed in the open position. The failure resulted from inappropriate work instructions for replacing the actuator diaphragm on the emergency feedwater to Steam Generator 1 backup isolation valve. As a result, the diaphragm was installed incorrectly, resulting in the failure on June 14, 2004. The finding was greater than minor because it affected the operability of a containment isolation valve and the availability of the emergency feedwater system, a mitigating system. The finding was of very low safety significance because a second isolation valve was available and could have performed the isolation function. The valve was promptly repaired and a condition report was initiated. The emergency feedwater system was inoperable for less than the allowed Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the failure to take adequate corrective action to ensure that the torque applied to the flow control valve for Accumulator B of main feedwater isolation Valve No. 1 was sufficient to prevent an O-ring from extruding, resulting in a loss of system hydraulic fluid and rendering the valve inoperable on June 20, 2004. The primary cause of the finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because it affected the reactor safety barrier cornerstone attribute for maintaining functionality of the containment boundary. The main feedwater isolation valve was repaired within the Technical Specification allowed outage time and a condition report was initiated. This finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment and the main feedwater isolation valve was inoperable for less time than the allowed by the Technical Specification outage time.
 
3Q/2004 Inspection Findings - Waterford 3                                                                                              Page 4 of 6 Inspection Report# : 2004008(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases. This failure resulted in significantly underestimating the actual dose to operators. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope and was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as condition report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for the failure to promptly identify and correct a condition adverse to quality. Specifically, Entergy failed to replace known age-degraded O-rings affecting the main feedwater isolation valves in the year 2000 resulting in O-ring failure and inoperability of the Train A feedwater isolation valve on December 27, 2003.
The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Maintenance Instruction Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to establish appropriate instructions for corrective maintenance activities on Train A main feedwater isolation valve on December 27, 2003. This resulted in the failure to establish appropriate torque specifications to ensure adequate O-ring compression that ultimately led to an O-ring failure and inoperability of the isolation valve on January 3, 2004. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity
 
3Q/2004 Inspection Findings - Waterford 3                                                                                            Page 5 of 6 of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Controls of MSIVs The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing main steam isolation Valves 1 and 2. This performance deficiency contributed to both valves being declared inoperable due to system leaks creating a low pressure condition in the valve actuating systems. This finding is more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance of the functionality of containment. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2003007(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Prevent Recurrence of PWSCC of Alloy 600 Material The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions resulting in recurrences of pressure boundary leakage due to primary water stress corrosion cracking of Alloy 600 reactor coolant system nozzles. This finding was greater than minor because it affected the reactor safety barrier integrity cornerstone objective for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Manual Chapter 0609 Significance determination process Phase 1 Screening Worksheet this performance deficiency affected the reactor coolant system barrier function requiring a Phase 2 analysis. The results of the Phase 2 and 3 analysis determined that this finding was of very low safety significance based on the cracks being axial in nature (does not contribute substantially to a loss of coolant accident) and the leaks resulted in a build up of only minor boric acid residue indicative of only trace amounts of through wall leakage. The leak rates identified were well within the capacity of a single charging pump.
Inspection Report# : 2003007(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Oct 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Barricade a High Radiation Area The inspector identified a noncited violation of Technical Specification 6.12.1 because Entergy failed to barricade a high radiation area.
Specifically, on October 27, 2003, the inspector observed that the high radiation area rope barricading the regenitive heat exchanger room was stretched across the entrance way at a height of approximately 79 inches, which would not obstruct the entry of station workers. General area radiation levels within the room were as high as 420 millirem per hour. The finding is in Entergy's corrective action program as Condition Report CR-WF3-2003-03164. The finding is greater than minor because it affected the Occupational Radiation Safety cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation and the finding is associated with the cornerstone attribute (Program & Process). The finding involved an individual's potential for unplanned or unintended dose. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with ALARA planning or work controls, there was no overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003007(pdf)
 
3Q/2004 Inspection Findings - Waterford 3                                                                                              Page 6 of 6 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A May 21, 2004 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the licensee's process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2004006(pdf)
Last modified : December 29, 2004
 
4Q/2004 Inspection Findings - Waterford 3                                                                                              Page 1 of 6 Waterford 3 4Q/2004 Plant Inspection Findings Initiating Events Significance:        Jun 26, 2004 Identified By: NRC Item Type: FIN Finding Improper Maintenance Activities resulting in Plant Down Power A self-revealing finding was identified involving improper installation of an O-ring for Emergency Header Check Valve EH-1285. This resulted in an unisolable hydraulic fluid leak in the main turbine electro-hydraulic control system. Entergy elected to reduce reactor power to less than 20 percent and manually trip the main turbine on February 14, 2004. This self-revealing finding is greater than minor because it is associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. The human performance attribute was affected in that the performance deficiency resulted in a perturbation in plant stability by reducing reactor power to less than 20 percent. Although the unisolable hydraulic leak resulted in a plant transient, the finding is of very low safety significance because it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004003(pdf)
Mitigating Systems Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Actions to Prevent Recurrence of Main Feedwater Isolation Valve Hydraulic System Over-Pressure Conditions The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the main feedwater isolation valves. Specifically, on multiple occasions accumulator over-pressure conditions have occurred, resulting from degraded hydraulic fluid adversely affecting the hydraulic actuator pressure relief system. These over-pressure conditions potentially result in valve closure stroke times outside design basis values. The finding was greater than minor because it is associated with the mitigating systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for mitigating systems. The finding was determined to be of very low risk significance because the over-pressure conditions did not represent an actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Control Room Electrical Isaolation During Transfer to the Alternate Shutdown Panel The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix R, Section III.L.3, for the failure to provide electrical independence in the Waterford design that included a neutral (ground) wire that was not isolated from the control room during transfer to the alternative shutdown panel. Entergy initiated Condition Report WF3-2004-03541 to track the modification to isolate the neutral wire for the affected safe shutdown circuits. The modification will bring Waterford into compliance with Appendix R. This finding is greater than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and it has the potential to impact the mitigating systems cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. The violation is associated with degradation of a fire protection feature. Using Part 1 of the Inspection Manual Chapter 0609, fire protection Significance Determination Process Phase 1 Worksheet, the performance issue was determined to be in the postfire safe shutdown category. The degradation rating was low based on Entergy's determination that there were no existing conditions that would prevent the plant from achieving and maintaining a safe shutdown in the event of a control room fire, if the installed protective devices always operated within their designed tripping characteristics. Therefore, the finding screens as Green or of very low safety significance in the
 
4Q/2004 Inspection Findings - Waterford 3                                                                                              Page 2 of 6 Phase 1 Worksheet. This violation is being treated as a noncited violation consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2004005(pdf)
Significance: N/A Sep 27, 2004 Identified By: NRC Item Type: FIN Finding Failure to establish appropriate instructions and to accomplish those instructions for installation of the emergency diesel generator Train A The NRC performed this supplemental inspection to assess the Entergy Operations, Inc. evaluation associated with the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel oil line for the Train A emergency diesel generator in May 2003. This was a violation of 10 CFR Part 50, Appendix B, Criterion V. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test, which rendered the Train A emergency diesel generator inoperable.
The NRC concluded that Entergy Operations, Inc. performed thorough evaluations of the emergency diesel generator fuel oil line failure. The root causes of the finding were adequately defined and understood. The corrective actions resulting from the evaluations appropriately addressed the identified causes. The contributing causes for the two noncited violations identified during this inspection are consistent with the finding from the diesel fuel oil line failure, and the corrective actions are consistent with the ongoing corrective actions to improve maintenance work instructions. This included development of work instructions for new and the remaking of existing compression fittings, establishment of maintenance technician qualification requirements for compression fittings, and development of training on tube bending.
Inspection Report# : 2004008(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Environmental Qualification Records The inspectors identified a noncited violation of 10 CFR 50.49(j) for the failure to maintain an auditable record demonstrating that electric equipment important to safety is environmentally qualified for its intended application. Specifically, it was identified that nonconservative temperature profiles were utilized to calculate the qualified life of ASCO NP8300 series solenoid-operated valves. The finding was more than minor since if left uncorrected it would become a more significant safety concern. Specifically, the failure to maintain electrical equipment in an environmentally qualified configuration could adversely impact the ability of such mitigating equipment to perform its safety function during design-basis accident conditions. This finding was of very low safety significance since additional analysis demonstrated that affected electrical equipment currently installed in the plant was environmentally qualifiable. Therefore, this deficiency did not result in any loss of affected equipment safety function.
Inspection Report# : 2004004(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Safety Injection Sump Recirculation Piping The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the containment safety injection sump recirculation piping. This deficiency resulted in inappropriately maintaining a section of the piping void of water, potentially affecting the operability of the high-pressure safety injection and containment spray pumps during postulated design-basis accident conditions following a recirculation actuation signal. This finding was more than minor because it potentially affected the mitigating system cornerstone objective of ensuring the capability of the high-pressure safety injection and containment spray systems to perform their design-basis functions. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004004(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct instances where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions. In one case, the licensee failed to properly address system operability and, for a two-week period, actual valve operability was unknown. The issue was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was determined to be of very low risk
 
4Q/2004 Inspection Findings - Waterford 3                                                                                                Page 3 of 6 significance because each issue: was not a design or qualification deficiency; did not result in the loss of a safety system; did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours; and was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensee's corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Diesel Generator Fuel Oil Storage Requirements The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the emergency diesel generating (EDG) system fuel oil storage requirements. This failure affected the ability of each emergency diesel generator to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offsite power and a design-bases accident.
This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency ac power to respond to initiating events to prevent undesirable consequences. This finding was evaluated using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the mitigating systems cornerstone. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in a loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct EDG Loading and Fuel Oil Consumption Analysis Deifciencies The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee inappropriately closed a corrective action requiring revisions to the EDG Loading and Fuel Oil Consumption analysis. The failure to adequately complete this corrective action resulted in the failure to maintain design control of the EDG fuel oil storage inventory requirements to ensure a 7-day post accident fuel oil inventory. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency ac power to respond to initiating events to prevent undesirable consequences. This finding was evaluated using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the mitigating systems cornerstone. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in a loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004002(pdf)
Significance:        Jan 05, 2004 Identified By: NRC Item Type: VIO Violation Failure to establish appropriate instructions and implement those instructions Contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings, during the overhaul of Train A emergency diesel generator in May 2003, the licensee failed to establish adequate instructions to ensure proper installation of th fuel supply line of Train A emergency diesel generator. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test.
Inspection Report# : 2003007(pdf)
Inspection Report# : 2004008(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2004 Inspection Findings - Waterford 3                                                                                            Page 4 of 6 Failure to Establish Adequate Test Controls for Leak Testing Fluid Systems Outside Containment that Contain High Radioactive Fluid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing those portions of fluid systems outside containment that could contain highly radioactive fluid during a serious transient or accident. This performance deficiency could result in underestimating the leak rate of highly radioactive fluid into the reactor auxiliary building during accident conditions. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for barrier integrity. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment and it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2004005(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Pevent Recurrence of Main Steam Isolation Valve Failures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to determine the cause and preclude recurrence of main steam isolation solenoid-operated dump valve failures. This failure affected the primary containment isolation function for the main steam system isolation valves. The primary cause of this finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because if left uncorrected the finding could become a more safety significant concern. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time. The valve was repaired and returned to service.
Inspection Report# : 2004004(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions Affecting the Emergency Feedwater System A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified when the valve failed in the open position. The failure resulted from inappropriate work instructions for replacing the actuator diaphragm on the emergency feedwater to Steam Generator 1 backup isolation valve. As a result, the diaphragm was installed incorrectly, resulting in the failure on June 14, 2004. The finding was greater than minor because it affected the operability of a containment isolation valve and the availability of the emergency feedwater system, a mitigating system. The finding was of very low safety significance because a second isolation valve was available and could have performed the isolation function. The valve was promptly repaired and a condition report was initiated. The emergency feedwater system was inoperable for less than the allowed Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the failure to take adequate corrective action to ensure that the torque applied to the flow control valve for Accumulator B of main feedwater isolation Valve No. 1 was sufficient to prevent an O-ring from extruding, resulting in a loss of system hydraulic fluid and rendering the valve inoperable on June 20, 2004. The primary cause of the finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because it affected the reactor safety barrier cornerstone attribute for maintaining functionality of the containment boundary. The main feedwater isolation valve was repaired within the Technical Specification allowed outage time and a condition report was initiated. This finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment and the main feedwater isolation valve was inoperable for less time than the allowed by the Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis
 
4Q/2004 Inspection Findings - Waterford 3                                                                                              Page 5 of 6 The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases. This failure resulted in significantly underestimating the actual dose to operators. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope and was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as condition report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI was identified for the failure to promptly identify and correct a condition adverse to quality. Specifically, Entergy failed to replace known age-degraded O-rings affecting the main feedwater isolation valves in the year 2000 resulting in O-ring failure and inoperability of the Train A feedwater isolation valve on December 27, 2003.
The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
Significance:        Mar 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Maintenance Instruction Affecting Main Feedwater Isolation Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to establish appropriate instructions for corrective maintenance activities on Train A main feedwater isolation valve on December 27, 2003. This resulted in the failure to establish appropriate torque specifications to ensure adequate O-ring compression that ultimately led to an O-ring failure and inoperability of the isolation valve on January 3, 2004. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment; it did not result in an actual open pathway affecting the physical integrity of reactor containment; and the main feedwater isolation valves were inoperable for less time than the allowed Technical Specification outage time.
Inspection Report# : 2004002(pdf)
 
4Q/2004 Inspection Findings - Waterford 3                                                                                              Page 6 of 6 Emergency Preparedness Occupational Radiation Safety Significance:        Nov 11, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Technical Specification Violation for Failure to Follow Radiation Work Permit Requirements The inspector identified a self-revealing noncited violation of Technical Specification 6.8.1 because Entergy failed to follow radiation work permit requirements. On November 12, 2003, two individuals' faces became contaminated while performing maintenance on Steam Generator 2 manway studs. Personnel contamination monitors alarmed upon the exit of the individuals from the controlled access area. These alarms prompted Entergy to investigate the events and conclude that multiple violations of Radiation Work Permit 2003-1509, Task 3, occurred.
Specifically, workers did not: (1) wear face shields or power visors during stud work, (2) have constant radiation protection technician coverage, (3) wear telemetry electronic dosimeters and move them to the head, or (4) wear lapel air samplers. This finding was entered into Entergy's corrective action program.
This finding is greater than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective because not following radiation work permit requirements could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A May 21, 2004 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the licensee's process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2004006(pdf)
Last modified : March 09, 2005
 
1Q/2005 Inspection Findings - Waterford 3                                                                                              Page 1 of 6 Waterford 3 1Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 26, 2004 Identified By: NRC Item Type: FIN Finding Improper Maintenance Activities resulting in Plant Down Power A self-revealing finding was identified involving improper installation of an O-ring for Emergency Header Check Valve EH-1285. This resulted in an unisolable hydraulic fluid leak in the main turbine electro-hydraulic control system. Entergy elected to reduce reactor power to less than 20 percent and manually trip the main turbine on February 14, 2004. This self-revealing finding is greater than minor because it is associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. The human performance attribute was affected in that the performance deficiency resulted in a perturbation in plant stability by reducing reactor power to less than 20 percent. Although the unisolable hydraulic leak resulted in a plant transient, the finding is of very low safety significance because it did not increase the likelihood of a primary or secondary system loss-of-coolant accident initiator, did not contribute to the loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004003(pdf)
Mitigating Systems Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Actions to Prevent Recurrence of Main Feedwater Isolation Valve Hydraulic System Over-Pressure Conditions The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the main feedwater isolation valves. Specifically, on multiple occasions accumulator over-pressure conditions have occurred, resulting from degraded hydraulic fluid adversely affecting the hydraulic actuator pressure relief system. These over-pressure conditions potentially result in valve closure stroke times outside design basis values. The finding was greater than minor because it is associated with the mitigating systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for mitigating systems. The finding was determined to be of very low risk significance because the over-pressure conditions did not represent an actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Control Room Electrical Isaolation During Transfer to the Alternate Shutdown Panel The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix R, Section III.L.3, for the failure to provide electrical independence in the Waterford design that included a neutral (ground) wire that was not isolated from the control room during transfer to the alternative shutdown panel. Entergy initiated Condition Report WF3-2004-03541 to track the modification to isolate the neutral wire for the affected safe shutdown circuits. The modification will bring Waterford into compliance with Appendix R. This finding is greater than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and it has the potential to impact the mitigating systems cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. The violation is associated with degradation of a fire protection feature. Using Part 1 of the Inspection Manual Chapter 0609, fire protection Significance Determination Process Phase 1 Worksheet, the performance issue was determined to be in the postfire safe shutdown category. The degradation rating was low based on Entergy's determination that there were no existing conditions that would prevent the plant from achieving and maintaining a safe shutdown in the event of a control room fire, if the installed protective devices always operated within their designed tripping characteristics. Therefore, the finding screens as Green or of very low safety significance in the
 
1Q/2005 Inspection Findings - Waterford 3                                                                                              Page 2 of 6 Phase 1 Worksheet. This violation is being treated as a noncited violation consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2004005(pdf)
Significance: N/A Sep 27, 2004 Identified By: NRC Item Type: FIN Finding Failure to establish appropriate instructions and to accomplish those instructions for installation of the emergency diesel generator Train A The NRC performed this supplemental inspection to assess the Entergy Operations, Inc. evaluation associated with the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel oil line for the Train A emergency diesel generator in May 2003. This was a violation of 10 CFR Part 50, Appendix B, Criterion V. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test, which rendered the Train A emergency diesel generator inoperable.
The NRC concluded that Entergy Operations, Inc. performed thorough evaluations of the emergency diesel generator fuel oil line failure. The root causes of the finding were adequately defined and understood. The corrective actions resulting from the evaluations appropriately addressed the identified causes. The contributing causes for the two noncited violations identified during this inspection are consistent with the finding from the diesel fuel oil line failure, and the corrective actions are consistent with the ongoing corrective actions to improve maintenance work instructions. This included development of work instructions for new and the remaking of existing compression fittings, establishment of maintenance technician qualification requirements for compression fittings, and development of training on tube bending.
Inspection Report# : 2004008(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Environmental Qualification Records The inspectors identified a noncited violation of 10 CFR 50.49(j) for the failure to maintain an auditable record demonstrating that electric equipment important to safety is environmentally qualified for its intended application. Specifically, it was identified that nonconservative temperature profiles were utilized to calculate the qualified life of ASCO NP8300 series solenoid-operated valves. The finding was more than minor since if left uncorrected it would become a more significant safety concern. Specifically, the failure to maintain electrical equipment in an environmentally qualified configuration could adversely impact the ability of such mitigating equipment to perform its safety function during design-basis accident conditions. This finding was of very low safety significance since additional analysis demonstrated that affected electrical equipment currently installed in the plant was environmentally qualifiable. Therefore, this deficiency did not result in any loss of affected equipment safety function.
Inspection Report# : 2004004(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Safety Injection Sump Recirculation Piping The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the containment safety injection sump recirculation piping. This deficiency resulted in inappropriately maintaining a section of the piping void of water, potentially affecting the operability of the high-pressure safety injection and containment spray pumps during postulated design-basis accident conditions following a recirculation actuation signal. This finding was more than minor because it potentially affected the mitigating system cornerstone objective of ensuring the capability of the high-pressure safety injection and containment spray systems to perform their design-basis functions. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004004(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct instances where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions. In one case, the licensee failed to properly address system operability and, for a two-week period, actual valve operability was unknown. The issue was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was determined to be of very low risk
 
1Q/2005 Inspection Findings - Waterford 3                                                                                                Page 3 of 6 significance because each issue: was not a design or qualification deficiency; did not result in the loss of a safety system; did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours; and was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensee's corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Test Controls for Leak Testing Fluid Systems Outside Containment that Contain High Radioactive Fluid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing those portions of fluid systems outside containment that could contain highly radioactive fluid during a serious transient or accident. This performance deficiency could result in underestimating the leak rate of highly radioactive fluid into the reactor auxiliary building during accident conditions. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for barrier integrity. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment and it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2004005(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Pevent Recurrence of Main Steam Isolation Valve Failures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to determine the cause and preclude recurrence of main steam isolation solenoid-operated dump valve failures. This failure affected the primary containment isolation function for the main steam system isolation valves. The primary cause of this finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because if left uncorrected the finding could become a more safety significant concern. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time. The valve was repaired and returned to service.
Inspection Report# : 2004004(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions Affecting the Emergency Feedwater System A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified when the valve failed in the open position. The failure resulted from inappropriate work instructions for replacing the actuator diaphragm on the emergency feedwater to Steam Generator 1 backup isolation valve. As a result, the diaphragm was installed incorrectly, resulting in the failure on June 14, 2004. The finding was greater than minor because it affected the operability of a containment isolation valve and the availability of the emergency feedwater system, a mitigating system. The finding was of very low safety significance because a second isolation valve was available and could have performed the isolation function. The valve was promptly repaired and a condition report was initiated. The emergency feedwater system was inoperable for less than the allowed Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        Aug 27, 2004
 
1Q/2005 Inspection Findings - Waterford 3                                                                                              Page 4 of 6 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the failure to take adequate corrective action to ensure that the torque applied to the flow control valve for Accumulator B of main feedwater isolation Valve No. 1 was sufficient to prevent an O-ring from extruding, resulting in a loss of system hydraulic fluid and rendering the valve inoperable on June 20, 2004. The primary cause of the finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because it affected the reactor safety barrier cornerstone attribute for maintaining functionality of the containment boundary. The main feedwater isolation valve was repaired within the Technical Specification allowed outage time and a condition report was initiated. This finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment and the main feedwater isolation valve was inoperable for less time than the allowed by the Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases. This failure resulted in significantly underestimating the actual dose to operators. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope and was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Significance:        May 21, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensee's corrective action program as condition report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2004006(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Nov 11, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Technical Specification Violation for Failure to Follow Radiation Work Permit Requirements The inspector identified a self-revealing noncited violation of Technical Specification 6.8.1 because Entergy failed to follow radiation work
 
1Q/2005 Inspection Findings - Waterford 3                                                                                              Page 5 of 6 permit requirements. On November 12, 2003, two individuals' faces became contaminated while performing maintenance on Steam Generator 2 manway studs. Personnel contamination monitors alarmed upon the exit of the individuals from the controlled access area. These alarms prompted Entergy to investigate the events and conclude that multiple violations of Radiation Work Permit 2003-1509, Task 3, occurred.
Specifically, workers did not: (1) wear face shields or power visors during stud work, (2) have constant radiation protection technician coverage, (3) wear telemetry electronic dosimeters and move them to the head, or (4) wear lapel air samplers. This finding was entered into Entergy's corrective action program.
This finding is greater than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective because not following radiation work permit requirements could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Significance:        Mar 04, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to ship radioactive material correctly The team reviewed a self-revealing, noncited violation of 10 CFR 71.5, which occurred when the licensee failed to ship radioactive material correctly. A radioactive shipment classified as an "excepted package-limited quantity" exceeded the external dose rate limitation of 0.5 millirem per hour on the surface of the package. The package recipient identified dose rates of 1.2 millirems per hour on the exterior surface of the package and notified the licensee of the problem. The finding is greater than minor because it was associated with a Public Radiation Safety cornerstone attribute (human performance) and it affected the associated cornerstone objective because the failure to correctly ship radioactive material decreases the licensee's assurance that the public will not receive unnecessary dose. However, this finding cannot be evaluated by the Public Radiation Safety Significance Determination Process because it did not involve radioactive shipments classified as Schedule 5 through 11, as described in NUREG-1660, and it did not fit traditional enforcement. Therefore, the finding was reviewed by NRC management and determined to be of very low safety significance. Additionally, this finding had cross-cutting aspects associated with human performance.
Licensee personnel directly contributed to the finding when they failed to ensure that the package did not exceed the dose rate limit. The finding was placed into the licensee's corrective action program as Condition Report WF3-2003-03514.
Inspection Report# : 2005009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A May 21, 2004 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the licensee's process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2004006(pdf)
 
1Q/2005 Inspection Findings - Waterford 3 Page 6 of 6 Last modified : June 17, 2005
 
2Q/2005 Inspection Findings - Waterford 3                                                                                                Page 1 of 6 Waterford 3 2Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Jun 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Train B Emergency Diesel Fuel Oil Storage Tank Level Instrument Sensing Line The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the Train B emergency diesel fuel oil storage tank level instrument sensing line resulting in level indication error. This error affected the ability of Train B fuel oil storage tank to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offiste power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding represented an actual loss of safety function, for a single train, for greater than its Technical Specification-allowed outage time, the finding was analyzed using Phase 2 of the Significant Determination Process. The finding was of very low safety significance because the licensee staff would have sufficient time to order replacement fuel, procedures existed to order replacement fuel and training was conducted on the existing procedures under conditions similar to the initiating event assumed.
Inspection Report# : 2005003(pdf)
Significance:        Apr 07, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent a Reoccurrence Cycle Timer Failure in the Essential Chiller The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the essential chillers. Specifically, on multiple occasions the essential chillers have failed to function as required due to cycle timer switch failure. Essential chiller malfunction could result in elevated chilled water system temperature used to cool areas containing safety significant equipment. This finding was more than minor in significance because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events and would become a more significant condition if left uncorrected. The inspectors utilized NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process Phase 1 Screening Worksheet, dated December 1, 2004, for Initiating Events, Mitigating Systems, and Barrier Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance because, for each essential chiller malfunction, the affected train was inoperable for less than the Technical Specification allowed outage time. A problem identification and resolution crosscutting aspect was identified for the failure to correct the condition which resulted in multiple timer failures (Section 4OA2).
Inspection Report# : 2005002(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Degraded performance could be masked and appropriate corrective actions not identified or implemented.
A finding with two examples of very low safety significance was identified for weaknesses in the maintenance rule program in regards to the component cooling water pumps, the reactor protection system and the reactor trip breakers. Specifically, the team found that the licensee did not monitor the performance or condition of structures, systems, or components in a manner sufficient to provide reasonable assurance that equipment reliability and degraded performance would not be masked and appropriate corrective actions would not be identified or implemented. This finding is more than minor because it affects the Mitigating Systems Cornerstone attributes of equipment reliability, in that, degraded performance could be masked and appropriate corrective actions not identified or implemented. This finding was of very low safety significance because no performance criteria were exceeded and there was no actual loss-of-safety function. Licensee personnel initiated Condition Report CR-WF3-2005-00322 to address this finding. (Section 1R21.4b2)
Inspection Report# : 2005008(pdf)
 
2Q/2005 Inspection Findings - Waterford 3                                                                                              Page 2 of 6 Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Failure to analyze the Dry Cooling Tower diesel driven sump pump discharge hose supports.
A finding of very low safety significance was identified for inadequate design of the diesel-driven sump pump associated with the dry cooling tower in that it did not provide an analysis to ensure that the support arrangement of the discharge hoses was adequate to support the discharge line. This finding is important to safety but not covered under 10 CFR Part 50, Appendix B Criterion. This finding was entered this issue into their corrective action program as Condition Report CR-WF3-2005-00592. This finding is greater than minor because it affected an attribute and the objective of the Mitigating Systems Cornerstone in that the design inadequacies did not provide assurance that the support arrangement for the diesel-driven sump pump was structurally adequate. The finding is of very low safety significance because, although it represented a design inadequacy, it did not contribute to a loss-of-mitigation equipment function, and did not increase the likelihood of a flood.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to develop and maintain procedures affecting a design basis limit.
A noncited violation of Waterford Technical Specification 6.8.1 was identified for failure to properly develop and implement procedures.
Technical Specification 6.8.1 states in part that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, which references activities affecting safety-related structures.
Contrary to this, station personnel failed to develop and implement procedures to relate the design basis ambient conditions to the operation of the ultimate heat sink cooling tower fans. As a result, no monitoring to recognize that a design basis limit has been exceeded, nor any actions required in the event that the design basis limit has been exceeded have been included in station procedures. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-0000590. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, if left uncorrected could result in the plant operating outside the design basis limits. The team determined this finding to be of very low safety significance because there was no evidence found that the licensee had exceeded their design basis limit.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control over Seismic Category 1 structure.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to perform a complete and adequate design of a Seismic Category 1 structure. Specifically, the licensee failed to perform a complete analysis of the component cooling water surge tank baffle plate. The surge tank was designed and constructed with a baffle plate internal to the tank, providing two independent trains of component cooling water. The analysis performed on the tank did not include an analysis of the baffle plate welds to ensure adequate performance for all applicable load scenarios. The licensee subsequently performed an analysis to demonstrate the adequacy of the baffle plate welds. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-00313. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, not providing adequate design analyses for the baffle plate welds did not ensure that all load scenarios were included in the analysis. Failure of these baffle plate welds could have resulted in a loss of both trains of component cooling water surge tank. This finding is determined to be of very low safety significance because the licensee performed a calculation that demonstrated the adequacy of the welds, and there was no actual loss of a safety function.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately analyze potential for over pressurizing ASME VIII air accumulators A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to provide justification for not providing over-pressure protection to air accumulators servicing safety-related valves, in accordance with ASME Code, Section VIII, Division
: 1. ASME Code, Section VIII, Division 1, paragraph UG-125, states that all pressure vessels (i.e., air accumulators), irrespective of size or pressure, shall be provided with pressure relief devices to protect against excessive pressure and these devices must be installed so that they may not be readily rendered inoperable. The team identified that the air accumulators, as installed, did not have any unisolable pressure relieving devices, therefore, causing the potential to over-pressure the air accumulators, challenging their structural integrity. The licensee had not provided an engineering analysis or justification for omitting over-pressure protection. The licensee initiated Condition Report CR-W3-2005-00596 to address NRC operability concerns. The finding is greater than minor because it affects Mitigating Systems Cornerstone in that not providing a design analysis did not ensure adequate protection against excessive pressure in air accumulators. Failure of these air accumulators could have resulted in a loss of motive force to the valves during loss of instrument air. Using the Phase 1 worksheet in Manual
 
2Q/2005 Inspection Findings - Waterford 3                                                                                              Page 3 of 6 Chapter 0609 "Significance Determination Process," the finding was determined to have very low safety significance because the air accumulators were later found to have a maximum allowable working pressure greater than the highest pressure that could be achieved in the system; therefore, the structural integrity of the design would not be challenged.
Inspection Report# : 2005008(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Actions to Prevent Recurrence of Main Feedwater Isolation Valve Hydraulic System Over-Pressure Conditions The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the main feedwater isolation valves. Specifically, on multiple occasions accumulator over-pressure conditions have occurred, resulting from degraded hydraulic fluid adversely affecting the hydraulic actuator pressure relief system. These over-pressure conditions potentially result in valve closure stroke times outside design basis values. The finding was greater than minor because it is associated with the mitigating systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for mitigating systems. The finding was determined to be of very low risk significance because the over-pressure conditions did not represent an actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Control Room Electrical Isaolation During Transfer to the Alternate Shutdown Panel The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix R, Section III.L.3, for the failure to provide electrical independence in the Waterford design that included a neutral (ground) wire that was not isolated from the control room during transfer to the alternative shutdown panel. Entergy initiated Condition Report WF3-2004-03541 to track the modification to isolate the neutral wire for the affected safe shutdown circuits. The modification will bring Waterford into compliance with Appendix R. This finding is greater than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and it has the potential to impact the mitigating systems cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. The violation is associated with degradation of a fire protection feature. Using Part 1 of the Inspection Manual Chapter 0609, fire protection Significance Determination Process Phase 1 Worksheet, the performance issue was determined to be in the postfire safe shutdown category. The degradation rating was low based on Entergy's determination that there were no existing conditions that would prevent the plant from achieving and maintaining a safe shutdown in the event of a control room fire, if the installed protective devices always operated within their designed tripping characteristics. Therefore, the finding screens as Green or of very low safety significance in the Phase 1 Worksheet. This violation is being treated as a noncited violation consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2004005(pdf)
Significance: N/A Sep 27, 2004 Identified By: NRC Item Type: FIN Finding Failure to establish appropriate instructions and to accomplish those instructions for installation of the emergency diesel generator Train A The NRC performed this supplemental inspection to assess the Entergy Operations, Inc. evaluation associated with the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel oil line for the Train A emergency diesel generator in May 2003. This was a violation of 10 CFR Part 50, Appendix B, Criterion V. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test, which rendered the Train A emergency diesel generator inoperable.
The NRC concluded that Entergy Operations, Inc. performed thorough evaluations of the emergency diesel generator fuel oil line failure. The root causes of the finding were adequately defined and understood. The corrective actions resulting from the evaluations appropriately addressed the identified causes. The contributing causes for the two noncited violations identified during this inspection are consistent with the finding from the diesel fuel oil line failure, and the corrective actions are consistent with the ongoing corrective actions to improve maintenance work instructions. This included development of work instructions for new and the remaking of existing compression fittings, establishment of maintenance technician qualification requirements for compression fittings, and development of training on tube bending.
Inspection Report# : 2004008(pdf)
Significance:        Sep 26, 2004 Identified By: NRC
 
2Q/2005 Inspection Findings - Waterford 3                                                                                            Page 4 of 6 Item Type: NCV NonCited Violation Failure to Maintain Adequate Environmental Qualification Records The inspectors identified a noncited violation of 10 CFR 50.49(j) for the failure to maintain an auditable record demonstrating that electric equipment important to safety is environmentally qualified for its intended application. Specifically, it was identified that nonconservative temperature profiles were utilized to calculate the qualified life of ASCO NP8300 series solenoid-operated valves. The finding was more than minor since if left uncorrected it would become a more significant safety concern. Specifically, the failure to maintain electrical equipment in an environmentally qualified configuration could adversely impact the ability of such mitigating equipment to perform its safety function during design-basis accident conditions. This finding was of very low safety significance since additional analysis demonstrated that affected electrical equipment currently installed in the plant was environmentally qualifiable. Therefore, this deficiency did not result in any loss of affected equipment safety function.
Inspection Report# : 2004004(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of Safety Injection Sump Recirculation Piping The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the containment safety injection sump recirculation piping. This deficiency resulted in inappropriately maintaining a section of the piping void of water, potentially affecting the operability of the high-pressure safety injection and containment spray pumps during postulated design-basis accident conditions following a recirculation actuation signal. This finding was more than minor because it potentially affected the mitigating system cornerstone objective of ensuring the capability of the high-pressure safety injection and containment spray systems to perform their design-basis functions. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in loss of function per Generic Letter 91-18, Revision 1.
Inspection Report# : 2004004(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Test Controls for Leak Testing Fluid Systems Outside Containment that Contain High Radioactive Fluid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing those portions of fluid systems outside containment that could contain highly radioactive fluid during a serious transient or accident. This performance deficiency could result in underestimating the leak rate of highly radioactive fluid into the reactor auxiliary building during accident conditions. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for barrier integrity. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment and it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2004005(pdf)
Significance:        Sep 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Pevent Recurrence of Main Steam Isolation Valve Failures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to determine the cause and preclude recurrence of main steam isolation solenoid-operated dump valve failures. This failure affected the primary containment isolation function for the main steam system isolation valves. The primary cause of this finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because if left uncorrected the finding could become a more safety significant concern. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment, it did not result in an actual open pathway affecting the physical integrity of reactor containment, and the main steam isolation valves were inoperable for less time than the allowed Technical Specification outage time. The valve was repaired and returned to service.
Inspection Report# : 2004004(pdf)
 
2Q/2005 Inspection Findings - Waterford 3                                                                                            Page 5 of 6 Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions Affecting the Emergency Feedwater System A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified when the valve failed in the open position. The failure resulted from inappropriate work instructions for replacing the actuator diaphragm on the emergency feedwater to Steam Generator 1 backup isolation valve. As a result, the diaphragm was installed incorrectly, resulting in the failure on June 14, 2004. The finding was greater than minor because it affected the operability of a containment isolation valve and the availability of the emergency feedwater system, a mitigating system. The finding was of very low safety significance because a second isolation valve was available and could have performed the isolation function. The valve was promptly repaired and a condition report was initiated. The emergency feedwater system was inoperable for less than the allowed Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Significance:        Aug 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action Affecting Main Feedwater Isolation Valve A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the failure to take adequate corrective action to ensure that the torque applied to the flow control valve for Accumulator B of main feedwater isolation Valve No. 1 was sufficient to prevent an O-ring from extruding, resulting in a loss of system hydraulic fluid and rendering the valve inoperable on June 20, 2004. The primary cause of the finding was related to the crosscutting area of problem identification and resolution. The finding was greater than minor because it affected the reactor safety barrier cornerstone attribute for maintaining functionality of the containment boundary. The main feedwater isolation valve was repaired within the Technical Specification allowed outage time and a condition report was initiated. This finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment and the main feedwater isolation valve was inoperable for less time than the allowed by the Technical Specification outage time.
Inspection Report# : 2004008(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Nov 11, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Technical Specification Violation for Failure to Follow Radiation Work Permit Requirements The inspector identified a self-revealing noncited violation of Technical Specification 6.8.1 because Entergy failed to follow radiation work permit requirements. On November 12, 2003, two individuals' faces became contaminated while performing maintenance on Steam Generator 2 manway studs. Personnel contamination monitors alarmed upon the exit of the individuals from the controlled access area. These alarms prompted Entergy to investigate the events and conclude that multiple violations of Radiation Work Permit 2003-1509, Task 3, occurred.
Specifically, workers did not: (1) wear face shields or power visors during stud work, (2) have constant radiation protection technician coverage, (3) wear telemetry electronic dosimeters and move them to the head, or (4) wear lapel air samplers. This finding was entered into Entergy's corrective action program.
This finding is greater than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective because not following radiation work permit requirements could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.
Inspection Report# : 2004005(pdf)
Public Radiation Safety
 
2Q/2005 Inspection Findings - Waterford 3                                                                                          Page 6 of 6 Significance:      Mar 04, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to ship radioactive material correctly The team reviewed a self-revealing, noncited violation of 10 CFR 71.5, which occurred when the licensee failed to ship radioactive material correctly. A radioactive shipment classified as an "excepted package-limited quantity" exceeded the external dose rate limitation of 0.5 millirem per hour on the surface of the package. The package recipient identified dose rates of 1.2 millirems per hour on the exterior surface of the package and notified the licensee of the problem. The finding is greater than minor because it was associated with a Public Radiation Safety cornerstone attribute (human performance) and it affected the associated cornerstone objective because the failure to correctly ship radioactive material decreases the licensee's assurance that the public will not receive unnecessary dose. However, this finding cannot be evaluated by the Public Radiation Safety Significance Determination Process because it did not involve radioactive shipments classified as Schedule 5 through 11, as described in NUREG-1660, and it did not fit traditional enforcement. Therefore, the finding was reviewed by NRC management and determined to be of very low safety significance. Additionally, this finding had cross-cutting aspects associated with human performance.
Licensee personnel directly contributed to the finding when they failed to ensure that the package did not exceed the dose rate limit. The finding was placed into the licensee's corrective action program as Condition Report WF3-2003-03514.
Inspection Report# : 2005009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 24, 2005
 
3Q/2005 Inspection Findings - Waterford 3                                                                                                Page 1 of 7 Waterford 3 3Q/2005 Plant Inspection Findings Initiating Events Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation (1)Failue to implement required procdure for RCS draindown; (2) Failure to implement required procedure for peer-checking; and (3)
Failure to perform an adequate prejob brief prior to reducing level A self-revealing noncited violation with three examples of Technical Specification 6.8.1.a was identified. The first involved the failure to implement Procedure OP-001-003, "RCS Drain Down," in establishing a reactor coolant system vent path when a nuclear auxiliary operator failed to open the reactor vessel vent line isolation valve as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The second violation of Technical Specification 6.8.1.a involved the failure to implement Procedure OP-100-001, "Operation Standards and Management Expectations," for providing a proper peer check for valve manipulations when a nuclear auxiliary operator failed to provide the required local peer check for opening the reactor vessel vent line isolation valve and erroneously agreed with the report that the valve had been properly opened and a vessel head vent path had been established. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The third violation of Technical Specification 6.8.1.a was identified for failure of the prejob brief to provide the nuclear auxiliary operators the required knowledge and information needed to successfully establish vent paths for the pressurizer and reactor vessel as required by procedure. The nuclear auxiliary operators responsible for establishing the vent paths did not attend this briefing as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463.
This finding has human performance crosscutting aspects associated with three failures to follow procedure. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown operations. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Sections 3.3.1, 3.3.2, and 3.3.4).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for reactor coolant system draindown A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern reactor coolant system inventory reductions. The reactor coolant system draindown procedure failed to identify that temporary vent rigs, required by procedure to properly establish vent paths, included in-line ball valves in series with the vent path and also failed to direct that those ball valves be opened to establish the vent path. As a result of this procedural inadequacy, one of the vent rig ball valves remained closed and the reactor coolant system remained unvented during the subsequent draindown, which caused the pressure in the reactor coolant system to drop below atmospheric. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. This finding has problem identification and resolution crosscutting aspects because the licensee was aware of and did not fix the procedure to address the ball valves in 2002. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the procedure quality attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations.
This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.3.3).
Inspection Report# : 2005010(pdf)
 
3Q/2005 Inspection Findings - Waterford 3                                                                                              Page 2 of 7 Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for performing the containment integrated leak rate test A self-revealing, noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern the integrated leak rate test for the containment vessel. The procedure for the test failed to prevent a plant configuration that allowed air to be entrained in the reactor coolant system and subsequently come out of solution and form a void in the reactor vessel head. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-2461. This finding has a human performance crosscutting aspect associated with procedure quality. This finding is more than minor because it is associated with the configuration control attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspector utilized the NRC Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Screen Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance since adequate mitigation capability remained available (Section 3.13).
Inspection Report# : 2005010(pdf)
Mitigating Systems Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change to a Method of Evaluation Without Prior NRC Approval The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the failure to obtain NRC approval prior to implementing a change to the facility that resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases. Specifically, the licensee implemented a change that assumed the unprotected dry cooling towers would not be impacted during a tornado event. This change was implemented based on the inappropriate use of a Tornado Missile Risk Evaluation method of evaluation not previously approved by the NRC. The licensee implemented this change to compensate for a licensee identified analysis error that adversely affected the ultimate heat sink capability following a tornado event. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The finding is greater than minor in that it affected the mitigating systems cornerstone attribute of equipment availability and function during a design bases tornado event.
Regional and NRR staff determined that the change made by the licensee resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases and that the change would require NRC approval under 10 CFR 50.59 guidance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed directly through the significance determination process. Therefore, this issue was considered applicable as traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the technical result or condition of a 10 CFR 50.59 violation can be assessed through the significance determination process. The inspectors and the Region IV reactor analyst discussed the significance of this finding. A significance Determination Process Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of mitigating system safety function per Generic Letter 91-18 guidance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly identify and correct the vacuum condition in the reactor coolant system A self-revealing noncited violation of Criterion XVI of Appendix B of 10 CFR Part 50 was identified for the failure to promptly identify and correct the vacuum condition in the reactor coolant system during draindown, a condition adverse to quality. Control room operators missed several opportunities over a 32.5-hour period to identify that a vacuum had been drawn on the reactor coolant system to correct the vacuum condition. The licensee documented this issue and their corrective actions in Condition Report CR-WF3-2005-1463. This finding has crosscutting aspects associated with problem identification and resolution for the failure to promptly identify and correct the vacuum condition.
This finding is greater than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the mitigating systems cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.8).
 
3Q/2005 Inspection Findings - Waterford 3                                                                                                Page 3 of 7 Inspection Report# : 2005010(pdf)
Significance:        Jun 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Train B Emergency Diesel Fuel Oil Storage Tank Level Instrument Sensing Line The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the Train B emergency diesel fuel oil storage tank level instrument sensing line resulting in level indication error. This error affected the ability of Train B fuel oil storage tank to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offiste power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding represented an actual loss of safety function, for a single train, for greater than its Technical Specification-allowed outage time, the finding was analyzed using Phase 2 of the Significant Determination Process. The finding was of very low safety significance because the licensee staff would have sufficient time to order replacement fuel, procedures existed to order replacement fuel and training was conducted on the existing procedures under conditions similar to the initiating event assumed.
Inspection Report# : 2005003(pdf)
Significance:        Apr 07, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent a Reoccurrence Cycle Timer Failure in the Essential Chiller The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the essential chillers. Specifically, on multiple occasions the essential chillers have failed to function as required due to cycle timer switch failure. Essential chiller malfunction could result in elevated chilled water system temperature used to cool areas containing safety significant equipment. This finding was more than minor in significance because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events and would become a more significant condition if left uncorrected. The inspectors utilized NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process Phase 1 Screening Worksheet, dated December 1, 2004, for Initiating Events, Mitigating Systems, and Barrier Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance because, for each essential chiller malfunction, the affected train was inoperable for less than the Technical Specification allowed outage time. A problem identification and resolution crosscutting aspect was identified for the failure to correct the condition which resulted in multiple timer failures (Section 4OA2).
Inspection Report# : 2005002(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to develop and maintain procedures affecting a design basis limit.
A noncited violation of Waterford Technical Specification 6.8.1 was identified for failure to properly develop and implement procedures.
Technical Specification 6.8.1 states in part that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, which references activities affecting safety-related structures.
Contrary to this, station personnel failed to develop and implement procedures to relate the design basis ambient conditions to the operation of the ultimate heat sink cooling tower fans. As a result, no monitoring to recognize that a design basis limit has been exceeded, nor any actions required in the event that the design basis limit has been exceeded have been included in station procedures. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-0000590. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, if left uncorrected could result in the plant operating outside the design basis limits. The team determined this finding to be of very low safety significance because there was no evidence found that the licensee had exceeded their design basis limit.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Failure to analyze the Dry Cooling Tower diesel driven sump pump discharge hose supports.
A finding of very low safety significance was identified for inadequate design of the diesel-driven sump pump associated with the dry cooling tower in that it did not provide an analysis to ensure that the support arrangement of the discharge hoses was adequate to support the discharge line. This finding is important to safety but not covered under 10 CFR Part 50, Appendix B Criterion. This finding was entered this issue into their corrective action program as Condition Report CR-WF3-2005-00592. This finding is greater than minor because it affected an attribute
 
3Q/2005 Inspection Findings - Waterford 3                                                                                              Page 4 of 7 and the objective of the Mitigating Systems Cornerstone in that the design inadequacies did not provide assurance that the support arrangement for the diesel-driven sump pump was structurally adequate. The finding is of very low safety significance because, although it represented a design inadequacy, it did not contribute to a loss-of-mitigation equipment function, and did not increase the likelihood of a flood.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control over Seismic Category 1 structure.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to perform a complete and adequate design of a Seismic Category 1 structure. Specifically, the licensee failed to perform a complete analysis of the component cooling water surge tank baffle plate. The surge tank was designed and constructed with a baffle plate internal to the tank, providing two independent trains of component cooling water. The analysis performed on the tank did not include an analysis of the baffle plate welds to ensure adequate performance for all applicable load scenarios. The licensee subsequently performed an analysis to demonstrate the adequacy of the baffle plate welds. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-00313. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, not providing adequate design analyses for the baffle plate welds did not ensure that all load scenarios were included in the analysis. Failure of these baffle plate welds could have resulted in a loss of both trains of component cooling water surge tank. This finding is determined to be of very low safety significance because the licensee performed a calculation that demonstrated the adequacy of the welds, and there was no actual loss of a safety function.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Degraded performance could be masked and appropriate corrective actions not identified or implemented.
A finding with two examples of very low safety significance was identified for weaknesses in the maintenance rule program in regards to the component cooling water pumps, the reactor protection system and the reactor trip breakers. Specifically, the team found that the licensee did not monitor the performance or condition of structures, systems, or components in a manner sufficient to provide reasonable assurance that equipment reliability and degraded performance would not be masked and appropriate corrective actions would not be identified or implemented. This finding is more than minor because it affects the Mitigating Systems Cornerstone attributes of equipment reliability, in that, degraded performance could be masked and appropriate corrective actions not identified or implemented. This finding was of very low safety significance because no performance criteria were exceeded and there was no actual loss-of-safety function. Licensee personnel initiated Condition Report CR-WF3-2005-00322 to address this finding. (Section 1R21.4b2)
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately analyze potential for over pressurizing ASME VIII air accumulators A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to provide justification for not providing over-pressure protection to air accumulators servicing safety-related valves, in accordance with ASME Code, Section VIII, Division
: 1. ASME Code, Section VIII, Division 1, paragraph UG-125, states that all pressure vessels (i.e., air accumulators), irrespective of size or pressure, shall be provided with pressure relief devices to protect against excessive pressure and these devices must be installed so that they may not be readily rendered inoperable. The team identified that the air accumulators, as installed, did not have any unisolable pressure relieving devices, therefore, causing the potential to over-pressure the air accumulators, challenging their structural integrity. The licensee had not provided an engineering analysis or justification for omitting over-pressure protection. The licensee initiated Condition Report CR-W3-2005-00596 to address NRC operability concerns. The finding is greater than minor because it affects Mitigating Systems Cornerstone in that not providing a design analysis did not ensure adequate protection against excessive pressure in air accumulators. Failure of these air accumulators could have resulted in a loss of motive force to the valves during loss of instrument air. Using the Phase 1 worksheet in Manual Chapter 0609 "Significance Determination Process," the finding was determined to have very low safety significance because the air accumulators were later found to have a maximum allowable working pressure greater than the highest pressure that could be achieved in the system; therefore, the structural integrity of the design would not be challenged.
Inspection Report# : 2005008(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Actions to Prevent Recurrence of Main Feedwater Isolation Valve Hydraulic System Over-Pressure
 
3Q/2005 Inspection Findings - Waterford 3                                                                                              Page 5 of 7 Conditions The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the main feedwater isolation valves. Specifically, on multiple occasions accumulator over-pressure conditions have occurred, resulting from degraded hydraulic fluid adversely affecting the hydraulic actuator pressure relief system. These over-pressure conditions potentially result in valve closure stroke times outside design basis values. The finding was greater than minor because it is associated with the mitigating systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for mitigating systems. The finding was determined to be of very low risk significance because the over-pressure conditions did not represent an actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Control Room Electrical Isaolation During Transfer to the Alternate Shutdown Panel The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix R, Section III.L.3, for the failure to provide electrical independence in the Waterford design that included a neutral (ground) wire that was not isolated from the control room during transfer to the alternative shutdown panel. Entergy initiated Condition Report WF3-2004-03541 to track the modification to isolate the neutral wire for the affected safe shutdown circuits. The modification will bring Waterford into compliance with Appendix R. This finding is greater than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and it has the potential to impact the mitigating systems cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. The violation is associated with degradation of a fire protection feature. Using Part 1 of the Inspection Manual Chapter 0609, fire protection Significance Determination Process Phase 1 Worksheet, the performance issue was determined to be in the postfire safe shutdown category. The degradation rating was low based on Entergy's determination that there were no existing conditions that would prevent the plant from achieving and maintaining a safe shutdown in the event of a control room fire, if the installed protective devices always operated within their designed tripping characteristics. Therefore, the finding screens as Green or of very low safety significance in the Phase 1 Worksheet. This violation is being treated as a noncited violation consistent with Section VI.A of the Enforcement Policy.
Inspection Report# : 2004005(pdf)
Barrier Integrity Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Procedure for Containment Closure Following a Loss of Shutdown Cooling Event The inspectors identified a Green noncited violation of Technical Specification 6.8, "Procedures and Programs," for the failure to establish adequate procedures regarding containment closure following loss of shutdown cooling while in reduced reactor coolant inventory conditions.
This deficiency could result in loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The failure to establish adequate procedures for containment closure in reduced reactor coolant inventory conditions is greater than minor in that if left uncorrected the finding would become a more significant safety concern that could result in the loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the finding was assessed as a Type B finding. Through interviews and review of additional analysis the licensee provided reasonable assurance that following a loss of shutdown cooling containment closure would be performed prior to core uncovery with leakage less than 100 percent containment volume per day through the equipment hatch. Using MC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the licensee's three year rolling average collective dose was less than 135 person-rem. Based on these assessments the finding was determined to be of very low safety significance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Main Steam Line Through Wall Pipe Leakage The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to preclude recurrence of through wall pipe leakage on the main steam line pipe 2MS2-123. This deficiency resulted in an unisolable steam leak requiring NRC approval
 
3Q/2005 Inspection Findings - Waterford 3                                                                                              Page 6 of 7 to deviate from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N523-2 to perform temporary repairs preventing a plant shutdown. The licensee entered this deficiency into their corrective action program for resolution. The inspectors determined the cause of this finding was related to the problem identification and resolution crosscutting area. The finding is greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2005004(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Test Controls for Leak Testing Fluid Systems Outside Containment that Contain High Radioactive Fluid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Section XI, "Test Control," for the failure to establish adequate test controls for leak testing those portions of fluid systems outside containment that could contain highly radioactive fluid during a serious transient or accident. This performance deficiency could result in underestimating the leak rate of highly radioactive fluid into the reactor auxiliary building during accident conditions. The finding was greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was evaluated using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for barrier integrity. The finding was only of very low safety significance because it did not represent an actual reduction of the atmospheric pressure control function of the reactor containment and it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2004005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Nov 11, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Technical Specification Violation for Failure to Follow Radiation Work Permit Requirements The inspector identified a self-revealing noncited violation of Technical Specification 6.8.1 because Entergy failed to follow radiation work permit requirements. On November 12, 2003, two individuals' faces became contaminated while performing maintenance on Steam Generator 2 manway studs. Personnel contamination monitors alarmed upon the exit of the individuals from the controlled access area. These alarms prompted Entergy to investigate the events and conclude that multiple violations of Radiation Work Permit 2003-1509, Task 3, occurred.
Specifically, workers did not: (1) wear face shields or power visors during stud work, (2) have constant radiation protection technician coverage, (3) wear telemetry electronic dosimeters and move them to the head, or (4) wear lapel air samplers. This finding was entered into Entergy's corrective action program.
This finding is greater than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective because not following radiation work permit requirements could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Significance:        Mar 04, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ship radioactive material correctly
 
3Q/2005 Inspection Findings - Waterford 3                                                                                          Page 7 of 7 The team reviewed a self-revealing, noncited violation of 10 CFR 71.5, which occurred when the licensee failed to ship radioactive material correctly. A radioactive shipment classified as an "excepted package-limited quantity" exceeded the external dose rate limitation of 0.5 millirem per hour on the surface of the package. The package recipient identified dose rates of 1.2 millirems per hour on the exterior surface of the package and notified the licensee of the problem. The finding is greater than minor because it was associated with a Public Radiation Safety cornerstone attribute (human performance) and it affected the associated cornerstone objective because the failure to correctly ship radioactive material decreases the licensee's assurance that the public will not receive unnecessary dose. However, this finding cannot be evaluated by the Public Radiation Safety Significance Determination Process because it did not involve radioactive shipments classified as Schedule 5 through 11, as described in NUREG-1660, and it did not fit traditional enforcement. Therefore, the finding was reviewed by NRC management and determined to be of very low safety significance. Additionally, this finding had cross-cutting aspects associated with human performance.
Licensee personnel directly contributed to the finding when they failed to ensure that the package did not exceed the dose rate limit. The finding was placed into the licensee's corrective action program as Condition Report WF3-2003-03514.
Inspection Report# : 2005009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : November 30, 2005
 
4Q/2005 Inspection Findings - Waterford 3                                                                                                Page 1 of 6 Waterford 3 4Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Coolant System Leakage Detection System The inspectors identified a noncited violation for the failure to comply with Technical Specification 3.4.5, "Leakage Detection Systems" based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection System." Specifically, the containment fan cooler condensate flow switches were identified to not meet the the design requirements for detecting a one gallon per minute reactor coolant system leak. The licensee took prompt corrective actions and entered this issue into the corrective action program.
The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that the finding was of very low safety significance (Green) because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that the licensee missed several opportunities to identify this non-conforming condition (Section R22).
Inspection Report# : 2005005(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation (1)Failue to implement required procedure for RCS draindown; (2) Failure to implement required procedure for peer-checking; and (3) Failure to perform an adequate prejob brief prior to reducing level A self-revealing noncited violation with three examples of Technical Specification 6.8.1.a was identified. The first involved the failure to implement Procedure OP-001-003, "RCS Drain Down," in establishing a reactor coolant system vent path when a nuclear auxiliary operator failed to open the reactor vessel vent line isolation valve as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The second violation of Technical Specification 6.8.1.a involved the failure to implement Procedure OP-100-001, "Operation Standards and Management Expectations," for providing a proper peer check for valve manipulations when a nuclear auxiliary operator failed to provide the required local peer check for opening the reactor vessel vent line isolation valve and erroneously agreed with the report that the valve had been properly opened and a vessel head vent path had been established. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The third violation of Technical Specification 6.8.1.a was identified for failure of the prejob brief to provide the nuclear auxiliary operators the required knowledge and information needed to successfully establish vent paths for the pressurizer and reactor vessel as required by procedure. The nuclear auxiliary operators responsible for establishing the vent paths did not attend this briefing as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463.
This finding has human performance crosscutting aspects associated with three failures to follow procedure. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown operations. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Sections 3.3.1, 3.3.2, and 3.3.4).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for reactor coolant system draindown A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern
 
4Q/2005 Inspection Findings - Waterford 3                                                                                              Page 2 of 6 reactor coolant system inventory reductions. The reactor coolant system draindown procedure failed to identify that temporary vent rigs, required by procedure to properly establish vent paths, included in-line ball valves in series with the vent path and also failed to direct that those ball valves be opened to establish the vent path. As a result of this procedural inadequacy, one of the vent rig ball valves remained closed and the reactor coolant system remained unvented during the subsequent draindown, which caused the pressure in the reactor coolant system to drop below atmospheric. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. This finding has problem identification and resolution crosscutting aspects because the licensee was aware of and did not fix the procedure to address the ball valves in 2002. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the procedure quality attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations.
This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.3.3).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for performing the containment integrated leak rate test A self-revealing, noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern the integrated leak rate test for the containment vessel. The procedure for the test failed to prevent a plant configuration that allowed air to be entrained in the reactor coolant system and subsequently come out of solution and form a void in the reactor vessel head. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-2461. This finding has a human performance crosscutting aspect associated with procedure quality. This finding is more than minor because it is associated with the configuration control attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspector utilized the NRC Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Screen Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance since adequate mitigation capability remained available (Section 3.13).
Inspection Report# : 2005010(pdf)
Mitigating Systems Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change to a Method of Evaluation Without Prior NRC Approval The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the failure to obtain NRC approval prior to implementing a change to the facility that resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases. Specifically, the licensee implemented a change that assumed the unprotected dry cooling towers would not be impacted during a tornado event. This change was implemented based on the inappropriate use of a Tornado Missile Risk Evaluation method of evaluation not previously approved by the NRC. The licensee implemented this change to compensate for a licensee identified analysis error that adversely affected the ultimate heat sink capability following a tornado event. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The finding is greater than minor in that it affected the mitigating systems cornerstone attribute of equipment availability and function during a design bases tornado event.
Regional and NRR staff determined that the change made by the licensee resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases and that the change would require NRC approval under 10 CFR 50.59 guidance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed directly through the significance determination process. Therefore, this issue was considered applicable as traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the technical result or condition of a 10 CFR 50.59 violation can be assessed through the significance determination process. The inspectors and the Region IV reactor analyst discussed the significance of this finding. A significance Determination Process Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of mitigating system safety function per Generic Letter 91-18 guidance.
Inspection Report# : 2005004(pdf)
 
4Q/2005 Inspection Findings - Waterford 3                                                                                                Page 3 of 6 Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly identify and correct the vacuum condition in the reactor coolant system A self-revealing noncited violation of Criterion XVI of Appendix B of 10 CFR Part 50 was identified for the failure to promptly identify and correct the vacuum condition in the reactor coolant system during draindown, a condition adverse to quality. Control room operators missed several opportunities over a 32.5-hour period to identify that a vacuum had been drawn on the reactor coolant system to correct the vacuum condition. The licensee documented this issue and their corrective actions in Condition Report CR-WF3-2005-1463. This finding has crosscutting aspects associated with problem identification and resolution for the failure to promptly identify and correct the vacuum condition.
This finding is greater than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the mitigating systems cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.8).
Inspection Report# : 2005010(pdf)
Significance:        Jun 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Train B Emergency Diesel Fuel Oil Storage Tank Level Instrument Sensing Line The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the Train B emergency diesel fuel oil storage tank level instrument sensing line resulting in level indication error. This error affected the ability of Train B fuel oil storage tank to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offiste power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding represented an actual loss of safety function, for a single train, for greater than its Technical Specification-allowed outage time, the finding was analyzed using Phase 2 of the Significant Determination Process. The finding was of very low safety significance because the licensee staff would have sufficient time to order replacement fuel, procedures existed to order replacement fuel and training was conducted on the existing procedures under conditions similar to the initiating event assumed.
Inspection Report# : 2005003(pdf)
Significance:        Apr 07, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent a Reoccurrence Cycle Timer Failure in the Essential Chiller The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the essential chillers. Specifically, on multiple occasions the essential chillers have failed to function as required due to cycle timer switch failure. Essential chiller malfunction could result in elevated chilled water system temperature used to cool areas containing safety significant equipment. This finding was more than minor in significance because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events and would become a more significant condition if left uncorrected. The inspectors utilized NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process Phase 1 Screening Worksheet, dated December 1, 2004, for Initiating Events, Mitigating Systems, and Barrier Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance because, for each essential chiller malfunction, the affected train was inoperable for less than the Technical Specification allowed outage time. A problem identification and resolution crosscutting aspect was identified for the failure to correct the condition which resulted in multiple timer failures (Section 4OA2).
Inspection Report# : 2005002(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to develop and maintain procedures affecting a design basis limit.
A noncited violation of Waterford Technical Specification 6.8.1 was identified for failure to properly develop and implement procedures.
Technical Specification 6.8.1 states in part that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, which references activities affecting safety-related structures.
 
4Q/2005 Inspection Findings - Waterford 3                                                                                              Page 4 of 6 Contrary to this, station personnel failed to develop and implement procedures to relate the design basis ambient conditions to the operation of the ultimate heat sink cooling tower fans. As a result, no monitoring to recognize that a design basis limit has been exceeded, nor any actions required in the event that the design basis limit has been exceeded have been included in station procedures. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-0000590. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, if left uncorrected could result in the plant operating outside the design basis limits. The team determined this finding to be of very low safety significance because there was no evidence found that the licensee had exceeded their design basis limit.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Failure to analyze the Dry Cooling Tower diesel driven sump pump discharge hose supports.
A finding of very low safety significance was identified for inadequate design of the diesel-driven sump pump associated with the dry cooling tower in that it did not provide an analysis to ensure that the support arrangement of the discharge hoses was adequate to support the discharge line. This finding is important to safety but not covered under 10 CFR Part 50, Appendix B Criterion. This finding was entered this issue into their corrective action program as Condition Report CR-WF3-2005-00592. This finding is greater than minor because it affected an attribute and the objective of the Mitigating Systems Cornerstone in that the design inadequacies did not provide assurance that the support arrangement for the diesel-driven sump pump was structurally adequate. The finding is of very low safety significance because, although it represented a design inadequacy, it did not contribute to a loss-of-mitigation equipment function, and did not increase the likelihood of a flood.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control over Seismic Category 1 structure.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to perform a complete and adequate design of a Seismic Category 1 structure. Specifically, the licensee failed to perform a complete analysis of the component cooling water surge tank baffle plate. The surge tank was designed and constructed with a baffle plate internal to the tank, providing two independent trains of component cooling water. The analysis performed on the tank did not include an analysis of the baffle plate welds to ensure adequate performance for all applicable load scenarios. The licensee subsequently performed an analysis to demonstrate the adequacy of the baffle plate welds. This issue was entered into the corrective action program as Condition Report CR-WF3-2005-00313. The finding is greater than minor because it affects the Mitigating Systems Cornerstone objective, in that, not providing adequate design analyses for the baffle plate welds did not ensure that all load scenarios were included in the analysis. Failure of these baffle plate welds could have resulted in a loss of both trains of component cooling water surge tank. This finding is determined to be of very low safety significance because the licensee performed a calculation that demonstrated the adequacy of the welds, and there was no actual loss of a safety function.
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: FIN Finding Degraded performance could be masked and appropriate corrective actions not identified or implemented.
A finding with two examples of very low safety significance was identified for weaknesses in the maintenance rule program in regards to the component cooling water pumps, the reactor protection system and the reactor trip breakers. Specifically, the team found that the licensee did not monitor the performance or condition of structures, systems, or components in a manner sufficient to provide reasonable assurance that equipment reliability and degraded performance would not be masked and appropriate corrective actions would not be identified or implemented. This finding is more than minor because it affects the Mitigating Systems Cornerstone attributes of equipment reliability, in that, degraded performance could be masked and appropriate corrective actions not identified or implemented. This finding was of very low safety significance because no performance criteria were exceeded and there was no actual loss-of-safety function. Licensee personnel initiated Condition Report CR-WF3-2005-00322 to address this finding. (Section 1R21.4b2)
Inspection Report# : 2005008(pdf)
Significance:        Mar 10, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately analyze potential for over pressurizing ASME VIII air accumulators A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for failure to provide justification for not
 
4Q/2005 Inspection Findings - Waterford 3                                                                                              Page 5 of 6 providing over-pressure protection to air accumulators servicing safety-related valves, in accordance with ASME Code, Section VIII, Division
: 1. ASME Code, Section VIII, Division 1, paragraph UG-125, states that all pressure vessels (i.e., air accumulators), irrespective of size or pressure, shall be provided with pressure relief devices to protect against excessive pressure and these devices must be installed so that they may not be readily rendered inoperable. The team identified that the air accumulators, as installed, did not have any unisolable pressure relieving devices, therefore, causing the potential to over-pressure the air accumulators, challenging their structural integrity. The licensee had not provided an engineering analysis or justification for omitting over-pressure protection. The licensee initiated Condition Report CR-W3-2005-00596 to address NRC operability concerns. The finding is greater than minor because it affects Mitigating Systems Cornerstone in that not providing a design analysis did not ensure adequate protection against excessive pressure in air accumulators. Failure of these air accumulators could have resulted in a loss of motive force to the valves during loss of instrument air. Using the Phase 1 worksheet in Manual Chapter 0609 "Significance Determination Process," the finding was determined to have very low safety significance because the air accumulators were later found to have a maximum allowable working pressure greater than the highest pressure that could be achieved in the system; therefore, the structural integrity of the design would not be challenged.
Inspection Report# : 2005008(pdf)
Barrier Integrity Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Procedure for Containment Closure Following a Loss of Shutdown Cooling Event The inspectors identified a Green noncited violation of Technical Specification 6.8, "Procedures and Programs," for the failure to establish adequate procedures regarding containment closure following loss of shutdown cooling while in reduced reactor coolant inventory conditions.
This deficiency could result in loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The failure to establish adequate procedures for containment closure in reduced reactor coolant inventory conditions is greater than minor in that if left uncorrected the finding would become a more significant safety concern that could result in the loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the finding was assessed as a Type B finding. Through interviews and review of additional analysis the licensee provided reasonable assurance that following a loss of shutdown cooling containment closure would be performed prior to core uncovery with leakage less than 100 percent containment volume per day through the equipment hatch. Using MC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the licensee's three year rolling average collective dose was less than 135 person-rem. Based on these assessments the finding was determined to be of very low safety significance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Main Steam Line Through Wall Pipe Leakage The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to preclude recurrence of through wall pipe leakage on the main steam line pipe 2MS2-123. This deficiency resulted in an unisolable steam leak requiring NRC approval to deviate from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N523-2 to perform temporary repairs preventing a plant shutdown. The licensee entered this deficiency into their corrective action program for resolution. The inspectors determined the cause of this finding was related to the problem identification and resolution crosscutting area. The finding is greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2005004(pdf)
Emergency Preparedness Occupational Radiation Safety
 
4Q/2005 Inspection Findings - Waterford 3                                                                                          Page 6 of 6 Public Radiation Safety Significance:      Mar 04, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ship radioactive material correctly The team reviewed a self-revealing, noncited violation of 10 CFR 71.5, which occurred when the licensee failed to ship radioactive material correctly. A radioactive shipment classified as an "excepted package-limited quantity" exceeded the external dose rate limitation of 0.5 millirem per hour on the surface of the package. The package recipient identified dose rates of 1.2 millirems per hour on the exterior surface of the package and notified the licensee of the problem. The finding is greater than minor because it was associated with a Public Radiation Safety cornerstone attribute (human performance) and it affected the associated cornerstone objective because the failure to correctly ship radioactive material decreases the licensee's assurance that the public will not receive unnecessary dose. However, this finding cannot be evaluated by the Public Radiation Safety Significance Determination Process because it did not involve radioactive shipments classified as Schedule 5 through 11, as described in NUREG-1660, and it did not fit traditional enforcement. Therefore, the finding was reviewed by NRC management and determined to be of very low safety significance. Additionally, this finding had cross-cutting aspects associated with human performance.
Licensee personnel directly contributed to the finding when they failed to ensure that the package did not exceed the dose rate limit. The finding was placed into the licensee's corrective action program as Condition Report WF3-2003-03514.
Inspection Report# : 2005009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 03, 2006
 
1Q/2006 Inspection Findings - Waterford 3                                                                                                Page 1 of 4 Waterford 3 1Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Coolant System Leakage Detection System The inspectors identified a noncited violation for the failure to comply with Technical Specification 3.4.5, "Leakage Detection Systems" based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection System." Specifically, the containment fan cooler condensate flow switches were identified to not meet the the design requirements for detecting a one gallon per minute reactor coolant system leak. The licensee took prompt corrective actions and entered this issue into the corrective action program.
The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that the finding was of very low safety significance (Green) because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that the licensee missed several opportunities to identify this non-conforming condition (Section R22).
Inspection Report# : 2005005(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation (1)Failue to implement required procedure for RCS draindown; (2) Failure to implement required procedure for peer-checking; and (3) Failure to perform an adequate prejob brief prior to reducing level A self-revealing noncited violation with three examples of Technical Specification 6.8.1.a was identified. The first involved the failure to implement Procedure OP-001-003, "RCS Drain Down," in establishing a reactor coolant system vent path when a nuclear auxiliary operator failed to open the reactor vessel vent line isolation valve as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The second violation of Technical Specification 6.8.1.a involved the failure to implement Procedure OP-100-001, "Operation Standards and Management Expectations," for providing a proper peer check for valve manipulations when a nuclear auxiliary operator failed to provide the required local peer check for opening the reactor vessel vent line isolation valve and erroneously agreed with the report that the valve had been properly opened and a vessel head vent path had been established. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The third violation of Technical Specification 6.8.1.a was identified for failure of the prejob brief to provide the nuclear auxiliary operators the required knowledge and information needed to successfully establish vent paths for the pressurizer and reactor vessel as required by procedure. The nuclear auxiliary operators responsible for establishing the vent paths did not attend this briefing as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463.
This finding has human performance crosscutting aspects associated with three failures to follow procedure. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown operations. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Sections 3.3.1, 3.3.2, and 3.3.4).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for reactor coolant system draindown A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern
 
1Q/2006 Inspection Findings - Waterford 3                                                                                              Page 2 of 4 reactor coolant system inventory reductions. The reactor coolant system draindown procedure failed to identify that temporary vent rigs, required by procedure to properly establish vent paths, included in-line ball valves in series with the vent path and also failed to direct that those ball valves be opened to establish the vent path. As a result of this procedural inadequacy, one of the vent rig ball valves remained closed and the reactor coolant system remained unvented during the subsequent draindown, which caused the pressure in the reactor coolant system to drop below atmospheric. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. This finding has problem identification and resolution crosscutting aspects because the licensee was aware of and did not fix the procedure to address the ball valves in 2002. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the procedure quality attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations.
This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.3.3).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for performing the containment integrated leak rate test A self-revealing, noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern the integrated leak rate test for the containment vessel. The procedure for the test failed to prevent a plant configuration that allowed air to be entrained in the reactor coolant system and subsequently come out of solution and form a void in the reactor vessel head. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-2461. This finding has a human performance crosscutting aspect associated with procedure quality. This finding is more than minor because it is associated with the configuration control attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspector utilized the NRC Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Screen Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance since adequate mitigation capability remained available (Section 3.13).
Inspection Report# : 2005010(pdf)
Mitigating Systems Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change to a Method of Evaluation Without Prior NRC Approval The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the failure to obtain NRC approval prior to implementing a change to the facility that resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases. Specifically, the licensee implemented a change that assumed the unprotected dry cooling towers would not be impacted during a tornado event. This change was implemented based on the inappropriate use of a Tornado Missile Risk Evaluation method of evaluation not previously approved by the NRC. The licensee implemented this change to compensate for a licensee identified analysis error that adversely affected the ultimate heat sink capability following a tornado event. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The finding is greater than minor in that it affected the mitigating systems cornerstone attribute of equipment availability and function during a design bases tornado event.
Regional and NRR staff determined that the change made by the licensee resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases and that the change would require NRC approval under 10 CFR 50.59 guidance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed directly through the significance determination process. Therefore, this issue was considered applicable as traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the technical result or condition of a 10 CFR 50.59 violation can be assessed through the significance determination process. The inspectors and the Region IV reactor analyst discussed the significance of this finding. A significance Determination Process Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of mitigating system safety function per Generic Letter 91-18 guidance.
Inspection Report# : 2005004(pdf)
 
1Q/2006 Inspection Findings - Waterford 3                                                                                                Page 3 of 4 Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly identify and correct the vacuum condition in the reactor coolant system A self-revealing noncited violation of Criterion XVI of Appendix B of 10 CFR Part 50 was identified for the failure to promptly identify and correct the vacuum condition in the reactor coolant system during draindown, a condition adverse to quality. Control room operators missed several opportunities over a 32.5-hour period to identify that a vacuum had been drawn on the reactor coolant system to correct the vacuum condition. The licensee documented this issue and their corrective actions in Condition Report CR-WF3-2005-1463. This finding has crosscutting aspects associated with problem identification and resolution for the failure to promptly identify and correct the vacuum condition.
This finding is greater than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the mitigating systems cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.8).
Inspection Report# : 2005010(pdf)
Significance:        Jun 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of the Train B Emergency Diesel Fuel Oil Storage Tank Level Instrument Sensing Line The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, for the failure to maintain design control of the Train B emergency diesel fuel oil storage tank level instrument sensing line resulting in level indication error. This error affected the ability of Train B fuel oil storage tank to provide sufficient fuel oil to support 7 days of continuous diesel generator operations following a loss of offiste power and a design-bases accident. This finding was greater than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding represented an actual loss of safety function, for a single train, for greater than its Technical Specification-allowed outage time, the finding was analyzed using Phase 2 of the Significant Determination Process. The finding was of very low safety significance because the licensee staff would have sufficient time to order replacement fuel, procedures existed to order replacement fuel and training was conducted on the existing procedures under conditions similar to the initiating event assumed.
Inspection Report# : 2005003(pdf)
Significance:        Apr 07, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent a Reoccurrence Cycle Timer Failure in the Essential Chiller The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to implement effective corrective actions to prevent recurrence for a significant condition adverse to quality affecting operability of the essential chillers. Specifically, on multiple occasions the essential chillers have failed to function as required due to cycle timer switch failure. Essential chiller malfunction could result in elevated chilled water system temperature used to cool areas containing safety significant equipment. This finding was more than minor in significance because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events and would become a more significant condition if left uncorrected. The inspectors utilized NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process Phase 1 Screening Worksheet, dated December 1, 2004, for Initiating Events, Mitigating Systems, and Barrier Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance because, for each essential chiller malfunction, the affected train was inoperable for less than the Technical Specification allowed outage time. A problem identification and resolution crosscutting aspect was identified for the failure to correct the condition which resulted in multiple timer failures (Section 4OA2).
Inspection Report# : 2005002(pdf)
Barrier Integrity Significance:        Sep 26, 2005
 
1Q/2006 Inspection Findings - Waterford 3                                                                                              Page 4 of 4 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Procedure for Containment Closure Following a Loss of Shutdown Cooling Event The inspectors identified a Green noncited violation of Technical Specification 6.8, "Procedures and Programs," for the failure to establish adequate procedures regarding containment closure following loss of shutdown cooling while in reduced reactor coolant inventory conditions.
This deficiency could result in loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The failure to establish adequate procedures for containment closure in reduced reactor coolant inventory conditions is greater than minor in that if left uncorrected the finding would become a more significant safety concern that could result in the loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the finding was assessed as a Type B finding. Through interviews and review of additional analysis the licensee provided reasonable assurance that following a loss of shutdown cooling containment closure would be performed prior to core uncovery with leakage less than 100 percent containment volume per day through the equipment hatch. Using MC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the licensee's three year rolling average collective dose was less than 135 person-rem. Based on these assessments the finding was determined to be of very low safety significance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Main Steam Line Through Wall Pipe Leakage The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to preclude recurrence of through wall pipe leakage on the main steam line pipe 2MS2-123. This deficiency resulted in an unisolable steam leak requiring NRC approval to deviate from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N523-2 to perform temporary repairs preventing a plant shutdown. The licensee entered this deficiency into their corrective action program for resolution. The inspectors determined the cause of this finding was related to the problem identification and resolution crosscutting area. The finding is greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2005004(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : May 25, 2006
 
2Q/2006 Inspection Findings - Waterford 3                                                                                                    Page 1 of 4 Waterford 3 2Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Coolant System Leakage Detection System The inspectors identified a noncited violation for the failure to comply with Technical Specification 3.4.5, "Leakage Detection Systems" based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection System." Specifically, the containment fan cooler condensate flow switches were identified to not meet the the design requirements for detecting a one gallon per minute reactor coolant system leak. The licensee took prompt corrective actions and entered this issue into the corrective action program.
The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that the finding was of very low safety significance (Green) because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that the licensee missed several opportunities to identify this non-conforming condition (Section R22).
Inspection Report# : 2005005(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation (1)Failue to implement required procedure for RCS draindown; (2) Failure to implement required procedure for peer-checking; and (3)
Failure to perform an adequate prejob brief prior to reducing level A self-revealing noncited violation with three examples of Technical Specification 6.8.1.a was identified. The first involved the failure to implement Procedure OP-001-003, "RCS Drain Down," in establishing a reactor coolant system vent path when a nuclear auxiliary operator failed to open the reactor vessel vent line isolation valve as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The second violation of Technical Specification 6.8.1.a involved the failure to implement Procedure OP-100-001, "Operation Standards and Management Expectations," for providing a proper peer check for valve manipulations when a nuclear auxiliary operator failed to provide the required local peer check for opening the reactor vessel vent line isolation valve and erroneously agreed with the report that the valve had been properly opened and a vessel head vent path had been established. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. The third violation of Technical Specification 6.8.1.a was identified for failure of the prejob brief to provide the nuclear auxiliary operators the required knowledge and information needed to successfully establish vent paths for the pressurizer and reactor vessel as required by procedure. The nuclear auxiliary operators responsible for establishing the vent paths did not attend this briefing as required. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463. This finding has human performance crosscutting aspects associated with three failures to follow procedure. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown operations. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Sections 3.3.1, 3.3.2, and 3.3.4).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for reactor coolant system draindown A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern reactor coolant system inventory reductions. The reactor coolant system draindown procedure failed to identify that temporary vent rigs, required by
 
2Q/2006 Inspection Findings - Waterford 3                                                                                                    Page 2 of 4 procedure to properly establish vent paths, included in-line ball valves in series with the vent path and also failed to direct that those ball valves be opened to establish the vent path. As a result of this procedural inadequacy, one of the vent rig ball valves remained closed and the reactor coolant system remained unvented during the subsequent draindown, which caused the pressure in the reactor coolant system to drop below atmospheric. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-1463.
This finding has problem identification and resolution crosscutting aspects because the licensee was aware of and did not fix the procedure to address the ball valves in 2002. This finding is more than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the procedure quality attribute of the initiating events cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.3.3).
Inspection Report# : 2005010(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to establish an adequate procedure for performing the containment integrated leak rate test A self-revealing, noncited violation of Technical Specification 6.8.1.a was identified for failure to establish an adequate procedure to govern the integrated leak rate test for the containment vessel. The procedure for the test failed to prevent a plant configuration that allowed air to be entrained in the reactor coolant system and subsequently come out of solution and form a void in the reactor vessel head. The licensee documented this issue and its corrective actions in Condition Report CR-WF3-2005-2461. This finding has a human performance crosscutting aspect associated with procedure quality. This finding is more than minor because it is associated with the configuration control attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspector utilized the NRC Inspection Manual Chapter 0609 Significance Determination Process Phase 1 Screen Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones to assess the safety significance. The finding was determined to be of very low risk significance since adequate mitigation capability remained available (Section 3.13).
Inspection Report# : 2005010(pdf)
Mitigating Systems Significance: SL-IV Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change to a Method of Evaluation Without Prior NRC Approval The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the failure to obtain NRC approval prior to implementing a change to the facility that resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases. Specifically, the licensee implemented a change that assumed the unprotected dry cooling towers would not be impacted during a tornado event. This change was implemented based on the inappropriate use of a Tornado Missile Risk Evaluation method of evaluation not previously approved by the NRC. The licensee implemented this change to compensate for a licensee identified analysis error that adversely affected the ultimate heat sink capability following a tornado event. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The finding is greater than minor in that it affected the mitigating systems cornerstone attribute of equipment availability and function during a design bases tornado event. Regional and NRR staff determined that the change made by the licensee resulted in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases and that the change would require NRC approval under 10 CFR 50.59 guidance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed directly through the significance determination process. Therefore, this issue was considered applicable as traditional enforcement. Although the significance determination process is not designed to assess significance of violations that potentially impact or impede the regulatory process, the technical result or condition of a 10 CFR 50.59 violation can be assessed through the significance determination process. The inspectors and the Region IV reactor analyst discussed the significance of this finding. A significance Determination Process Phase 1 screening was performed and the finding was determined to have very low safety significance because there was no actual loss of mitigating system safety function per Generic Letter 91-18 guidance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 14, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly identify and correct the vacuum condition in the reactor coolant system A self-revealing noncited violation of Criterion XVI of Appendix B of 10 CFR Part 50 was identified for the failure to promptly identify and correct
 
2Q/2006 Inspection Findings - Waterford 3                                                                                                    Page 3 of 4 the vacuum condition in the reactor coolant system during draindown, a condition adverse to quality. Control room operators missed several opportunities over a 32.5-hour period to identify that a vacuum had been drawn on the reactor coolant system to correct the vacuum condition. The licensee documented this issue and their corrective actions in Condition Report CR-WF3-2005-1463. This finding has crosscutting aspects associated with problem identification and resolution for the failure to promptly identify and correct the vacuum condition. This finding is greater than minor because if left uncorrected it could have become a more safety significant concern, it was associated with the human performance attribute of the mitigating systems cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated utilizing Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, "Shutdown Operations," Checklist 2. Using the Phase 1 guidelines, the inspectors determined that the finding increased the likelihood that a loss of decay heat removal would occur due to a decrease in the available net positive suction head available to the operating shutdown cooling pumps at the low reactor coolant system pressure. The inspectors determined the finding required a Phase 2 analysis and was sent to the regional Senior Reactor Analysts for risk quantification. The risk was determined to be of very low safety significance because, in this case, the reactor coolant system level was being administratively limited at a level where the system was not vulnerable to air binding the shutdown cooling pumps (Section 3.8).
Inspection Report# : 2005010(pdf)
Barrier Integrity Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Adequate Procedure for Containment Closure Following a Loss of Shutdown Cooling Event The inspectors identified a Green noncited violation of Technical Specification 6.8, "Procedures and Programs," for the failure to establish adequate procedures regarding containment closure following loss of shutdown cooling while in reduced reactor coolant inventory conditions. This deficiency could result in loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable. The licensee entered this deficiency into their corrective action program for resolution. The cause of this finding is related to the crosscutting element of human performance. The failure to establish adequate procedures for containment closure in reduced reactor coolant inventory conditions is greater than minor in that if left uncorrected the finding would become a more significant safety concern that could result in the loss of the containment barrier when called upon and the failure to maintain occupational radiation exposures as low as reasonably achievable.
Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the finding was assessed as a Type B finding. Through interviews and review of additional analysis the licensee provided reasonable assurance that following a loss of shutdown cooling containment closure would be performed prior to core uncovery with leakage less than 100 percent containment volume per day through the equipment hatch. Using MC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the licensee's three year rolling average collective dose was less than 135 person-rem. Based on these assessments the finding was determined to be of very low safety significance.
Inspection Report# : 2005004(pdf)
Significance:        Sep 26, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Main Steam Line Through Wall Pipe Leakage The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to preclude recurrence of through wall pipe leakage on the main steam line pipe 2MS2-123. This deficiency resulted in an unisolable steam leak requiring NRC approval to deviate from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N523-2 to perform temporary repairs preventing a plant shutdown. The licensee entered this deficiency into their corrective action program for resolution. The inspectors determined the cause of this finding was related to the problem identification and resolution crosscutting area. The finding is greater than minor because it affected the reactor safety barrier integrity cornerstone for providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it did not result in an actual open pathway affecting the physical integrity of reactor containment.
Inspection Report# : 2005004(pdf)
Emergency Preparedness Occupational Radiation Safety
 
2Q/2006 Inspection Findings - Waterford 3                                                                                                  Page 4 of 4 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Mar 24, 2006 Identified By: NRC Item Type: FIN Finding identification and Resolution of Problems The team reviewed approximately 237 corrective action program documents, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. Based on this review, the team found the licensees process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification of degraded conditions in the plant. The team reviewed corrective actions associated with these degraded conditions and design issues at Waterford Steam Electric Station, Unit 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford Steam Electric Station, Unit 3 based upon interviews conducted with plant personnel. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2006008(pdf)
Last modified : August 25, 2006
 
3Q/2006 Inspection Findings - Waterford 3                                                                            Page 1 of 3 Waterford 3 3Q/2006 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Coolant System Leakage Detection System The inspectors identified a noncited violation for the failure to comply with Technical Specification 3.4.5, "Leakage Detection Systems" based on a method of reactor coolant system leakage detection not meeting design standards of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection System." Specifically, the containment fan cooler condensate flow switches were identified to not meet the the design requirements for detecting a one gallon per minute reactor coolant system leak. The licensee took prompt corrective actions and entered this issue into the corrective action program.
The finding is greater than minor because it is associated with the Initiating Event cornerstone attribute of equipment performance and affects the associated cornerstone objective to limit the likelihood of an event that might upset plant stability and challenge critical safety functions in that a reactor coolant system leak could go undetected until it became more severe. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that the finding was of very low safety significance (Green) because other methods of reactor coolant system leakage detection were available, mitigating systems would not be affected, and it did not contribute to the likelihood of a reactor trip. The finding has a crosscutting element related to problem identification and resolution in that the licensee missed several opportunities to identify this non-conforming condition (Section R22).
Inspection Report# : 2005005(pdf)
Mitigating Systems Significance: SL-IV Aug 09, 2006 Identified By: NRC Item Type: VIO Violation Inaccurate Performance Indicator Information The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations. Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRCs recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green
 
3Q/2006 Inspection Findings - Waterford 3                                                                            Page 2 of 3 threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem identification and resolution crosscutting aspects, in that the implementation of the licensees Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate.
Inspection Report# : 2006009(pdf)
Significance:        Jul 07, 2006 Identified By: NRC Item Type: VIO Violation Untimely Actions to Reestablish Full Qualification of the Emergency Diesel Generator Starting Air System The inspectors identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." In September 2003, the NRC identified that the emergency diesel generator starting air system was incapable of supplying sufficient air to start its respective emergency diesel generator a minimum of five times without being recharged. To date, the licensee has failed to take appropriate corrective actions in a timely manner to correct this deficiency and restore compliance.
This finding is greater than minor because it affected the mitigating system cornerstone objective due to the degradation of the design-basis capability of the starting air system. This finding has a crosscutting aspect in the corrective action component of the problem identification and resolution area because the licensee failed to take actions to address safety issues in a timely manner, commensurate with its safety significance and complexity. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2006003(pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Mar 24, 2006
 
3Q/2006 Inspection Findings - Waterford 3                                                                            Page 3 of 3 Identified By: NRC Item Type: FIN Finding identification and Resolution of Problems The team reviewed approximately 237 corrective action program documents, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. Based on this review, the team found the licensees process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification of degraded conditions in the plant. The team reviewed corrective actions associated with these degraded conditions and design issues at Waterford Steam Electric Station, Unit 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford Steam Electric Station, Unit 3 based upon interviews conducted with plant personnel. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2006008(pdf)
Last modified : December 21, 2006
 
4Q/2006 Inspection Findings - Waterford 3                                                                              Page 1 of 3 Waterford 3 4Q/2006 Plant Inspection Findings Initiating Events Significance:      Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure for ESFAS Relay Replacement A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for an inadequate procedure that resulted in the unintentional actuation of five engineered safety features actuation system Train B relays and the loss of a 480 Vac motor control center. The 480 Vac motor control center provided power to the Train B pressurizer heaters and to the control element assembly motor generator Set B. Loss of the control element assembly motor generator increased the likelihood of a reactor trip. This finding is greater than minor because it affects the Initiating Event cornerstone objective procedure quality attribute to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. This finding was evaluated using the significance determination process and was determined to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a crosscutting aspect in the area of human performance associated with resources because the licensee failed to ensure that Work Order 26998 was adequate for the task.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:      Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Inspection of Essential Chiller Condenser Tubing A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failing to follow a maintenance procedure during performance of eddy current testing on the safety-related essential chiller Train A condenser tubing. The performance deficiency was the failure to perform a full length eddy current inspection of each tube with an appropriately sized eddy current probe. Subsequently, essential chiller Train A was removed from service to correct a throughwall tube leak in its condenser. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone because the performance deficiency affected the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Appendix A, Phase 1, questions for mitigating systems, the inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, there was no loss of a safety function, and there were no other adverse impacts to the facility. This finding had a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively communicate expectations of procedure compliance.
Inspection Report# : 2006004 (pdf)
Significance: SL-IV Aug 09, 2006 Identified By: NRC Item Type: VIO Violation Inaccurate Performance Indicator Information The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to
 
4Q/2006 Inspection Findings - Waterford 3                                                                            Page 2 of 3 the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations. Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRCs recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem identification and resolution crosscutting aspects, in that the implementation of the licensees Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate.
Inspection Report# : 2006009 (pdf)
Significance:        Jul 07, 2006 Identified By: NRC Item Type: VIO Violation Untimely Actions to Reestablish Full Qualification of the Emergency Diesel Generator Starting Air System The inspectors identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." In September 2003, the NRC identified that the emergency diesel generator starting air system was incapable of supplying sufficient air to start its respective emergency diesel generator a minimum of five times without being recharged. To date, the licensee has failed to take appropriate corrective actions in a timely manner to correct this deficiency and restore compliance.
This finding is greater than minor because it affected the mitigating system cornerstone objective due to the degradation of the design-basis capability of the starting air system. This finding has a crosscutting aspect in the corrective action component of the problem identification and resolution area because the licensee failed to take actions to address safety issues in a timely manner, commensurate with its safety significance and complexity. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2006003 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety
 
4Q/2006 Inspection Findings - Waterford 3                                                                            Page 3 of 3 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Mar 24, 2006 Identified By: NRC Item Type: FIN Finding identification and Resolution of Problems The team reviewed approximately 237 corrective action program documents, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. Based on this review, the team found the licensees process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification of degraded conditions in the plant. The team reviewed corrective actions associated with these degraded conditions and design issues at Waterford Steam Electric Station, Unit 3, which had cross-cutting aspects in the area of problem identification and resolution. The team concluded that a positive safety-conscience work environment exists at Waterford Steam Electric Station, Unit 3 based upon interviews conducted with plant personnel. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
Inspection Report# : 2006008 (pdf)
Last modified : March 01, 2007
 
Waterford 3 1Q/2007 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Excess Torque Resulting in Pressurizer Skirt Bolt Failures A self-revealing violation of very low safety significance of Technical Specification 6.8.1.a was identified for an inadequate procedure for installing a bolted joint that provided structural support for the pressurizer. Specifically, the installation procedure required applying 8750 ft-lbs torque to make up a bolted joint. Following corrective actions, the licensee discovered that the break away torque on several bolts exceeded 13,400 ft-lbs. The improper bolt tensioning resulted in failure of 1 of 16 bolts and the partial cracking of 3 other bolts that potentially could affect the pressurizers function in a safe shutdown earthquake event. This finding is more than minor because if left uncorrected it could have become a more safety significant concern. The finding was associated with the equipment performance attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. This finding was determined to have very low safety significance because a seismic event would not result in a loss-of-coolant accident that exceeded the Technical Specification limit for reactor coolant system leakage. Therefore, this issue screened out in Phase 1 of the MC 0609 significance determination because there was no actual loss of safety function.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure for ESFAS Relay Replacement A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for an inadequate procedure that resulted in the unintentional actuation of five engineered safety features actuation system Train B relays and the loss of a 480 Vac motor control center. The 480 Vac motor control center provided power to the Train B pressurizer heaters and to the control element assembly motor generator Set B. Loss of the control element assembly motor generator increased the likelihood of a reactor trip. This finding is greater than minor because it affects the Initiating Event cornerstone objective procedure quality attribute to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. This finding was evaluated using the significance determination process and was determined to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a crosscutting aspect in the area of human performance associated with resources because the licensee failed to ensure that Work Order 26998 was adequate for the task.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Recurring Failure of Valve SI-405B to Open A self-revealing violation of very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion XVI,
 
Corrective Action, was identified for the failure to implement effective corrective actions to prevent recurrence of a significant condition adverse to quality. Specifically, on multiple occasions Valve SI-405B failed to stroke open while attempting to place shutdown cooling Train B in service. This violation of Appendix B, Criterion XVI, is being treated as an noncited violation and was entered into the licensees corrective action program. This finding is greater than minor because it affects the the Mitigating Systems Cornerstone attribute of equipment operability, availability, and reliability of systems that respond to initiating events. This finding was evaluated using the significance determination process and was determined to be a finding of very low safety significance because, in each condition identified, it did not represent an actual loss of a safety function. The inspectors also determined that the cause of the condition had crosscutting aspects associated with the corrective action program component in the problem identification and resolution area. This assessment was based on the fact that the licensee failed to thoroughly evaluate the problem such that the resolutions addressed the causes and therefore, corrective actions were inadequate to prevent repetition.
Inspection Report# : 2006005 (pdf)
Significance:      Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Inspection of Essential Chiller Condenser Tubing A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failing to follow a maintenance procedure during performance of eddy current testing on the safety-related essential chiller Train A condenser tubing. The performance deficiency was the failure to perform a full length eddy current inspection of each tube with an appropriately sized eddy current probe. Subsequently, essential chiller Train A was removed from service to correct a throughwall tube leak in its condenser. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone because the performance deficiency affected the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Appendix A, Phase 1, questions for mitigating systems, the inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, there was no loss of a safety function, and there were no other adverse impacts to the facility. This finding had a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively communicate expectations of procedure compliance.
Inspection Report# : 2006004 (pdf)
Significance: SL-IV Aug 09, 2006 Identified By: NRC Item Type: VIO Violation Inaccurate Performance Indicator Information The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations. Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRCs recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem
 
identification and resolution crosscutting aspects, in that the implementation of the licensees Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate.
Inspection Report# : 2006009 (pdf)
Significance:        Jul 07, 2006 Identified By: NRC Item Type: VIO Violation Untimely Actions to Reestablish Full Qualification of the Emergency Diesel Generator Starting Air System The inspectors identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." In September 2003, the NRC identified that the emergency diesel generator starting air system was incapable of supplying sufficient air to start its respective emergency diesel generator a minimum of five times without being recharged. To date, the licensee has failed to take appropriate corrective actions in a timely manner to correct this deficiency and restore compliance.
This finding is greater than minor because it affected the mitigating system cornerstone objective due to the degradation of the design-basis capability of the starting air system. This finding has a crosscutting aspect in the corrective action component of the problem identification and resolution area because the licensee failed to take actions to address safety issues in a timely manner, commensurate with its safety significance and complexity. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2006003 (pdf)
Barrier Integrity Significance:        Feb 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct an Adverse Condition (Welds Not In Accordance With Design)
A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to promptly identify and correct an adverse condition (i.e., steam generator batwing-to-wrapper bar welds not in accordance with design).
Specifically, in May 2005, during Refueling Cycle 13, licensee personnel found that the batwing-to-wrapper bar welds were not in accordance with design drawings, but did not enter the adverse condition into the corrective action program until December 2006. This condition was entered into the corrective action program as Condition Report WF3-2006-04395. This finding was more than minor because by not promptly entering the non-conforming welds into the corrective action program and taking actions to correct the adverse condition, it became a more significant condition when two welds failed during Operating Cycle 14. Using the guidance of Appendix J to NRC Inspection Manual Chapter 0609, Significance Determination Process, the finding is determined to have very low safety significance (Green) because there was no tube degradation that exceeded 40 percent through-wall which did not increase in the large early release frequency.
This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action) program component.
Inspection Report# : 2006012 (pdf)
Emergency Preparedness Significance:        Dec 20, 2006 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Conduct a Required Offsite medical Drill in 2005 The inspector identified a noncited violation of 10 CFR 50.54(q) for failure to conduct during 2005 an offsite drill involving a simulated contaminated individual with provision for participation by local medical support services as required by the licensees emergency plan. The licensee failure to conduct the drill is a performance deficiency because the licensee identified the drills postponement in October 2005 and did not appropriately reschedule the drill. The licensee did not request NRC approval to deviate from this emergency plan requirement. This finding is greater than minor because a degraded proficiency in providing appropriate medical treatment for a contaminated individual has a potential impact on the safety of licensee employees and the public. The finding is of very low safety significance because the licensee failed to conduct only one required drill during the inspection period January 2005 through December 2006, and the drill was not appropriately rescheduled with NRC approval. This finding is a non-cited violation of 10 CFR 50.54(q) and 10 CFR 50 Appendix E, IV, F.1. The licensee has entered this issue into their corrective action system as Condition Report 2006-4429.
Inspection Report# : 2006005 (pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : June 01, 2007
 
Waterford 3 2Q/2007 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Excess Torque Resulting in Pressurizer Skirt Bolt Failures A self-revealing violation of very low safety significance of Technical Specification 6.8.1.a was identified for an inadequate procedure for installing a bolted joint that provided structural support for the pressurizer. Specifically, the installation procedure required applying 8750 ft-lbs torque to make up a bolted joint. Following corrective actions, the licensee discovered that the break away torque on several bolts exceeded 13,400 ft-lbs. The improper bolt tensioning resulted in failure of 1 of 16 bolts and the partial cracking of 3 other bolts that potentially could affect the pressurizers function in a safe shutdown earthquake event. This finding is more than minor because if left uncorrected it could have become a more safety significant concern. The finding was associated with the equipment performance attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations.
This finding was determined to have very low safety significance because a seismic event would not result in a loss-of-coolant accident that exceeded the Technical Specification limit for reactor coolant system leakage. Therefore, this issue screened out in Phase 1 of the MC 0609 significance determination because there was no actual loss of safety function.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure for ESFAS Relay Replacement A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for an inadequate procedure that resulted in the unintentional actuation of five engineered safety features actuation system Train B relays and the loss of a 480 Vac motor control center. The 480 Vac motor control center provided power to the Train B pressurizer heaters and to the control element assembly motor generator Set B. Loss of the control element assembly motor generator increased the likelihood of a reactor trip. This finding is greater than minor because it affects the Initiating Event cornerstone objective procedure quality attribute to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. This finding was evaluated using the significance determination process and was determined to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a crosscutting aspect in the area of human performance associated with resources because the licensee failed to ensure that Work Order 26998 was adequate for the task.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:        Apr 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Design Basis into Drawings
 
The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III, Design Control, for failure to assure that the design basis, as specified in the license application, was correctly translated into drawings and the actual plant configuration. Specifically, Waterford Final Safety Analysis Report, Section 2.4.2.3.3.d, describes openings in the dry cooling tower cubicles that help preclude the possibility of flooding Motor Control Centers 3A315-S and 3B315-S during the probable maximum precipitation event. These openings serve as a backup to the floor drains located in each cubicle. Current plant configuration and Drawing G-499 S06, Common Foundation Structure, Masonry, Sheet 6, do not conform to the design basis, in that there are no openings other than the floor drains. These motor control centers control power to the wet and dry cooling tower fans, which act as the ultimate heat sink. The licensee entered this issue into their corrective action program for resolution. This finding is more than minor because it is associated with the design control attribute and affects the Mitigating Systems cornerstone objective to ensure the reliability of the dry cooling tower system during the probable maximum precipitation event on the plant site. The normal floor drains had historically clogged and the drainage openings were needed to limit flood related challenges to the motor control centers. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the wet and dry cooling tower systems safety functions during the past year per Part 9900: Technical Guidance, Operability Determinations & Functional Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality.
Inspection Report# : 2007002 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: FIN Finding Failure to Ensure that Written Procedures Adequately Incorporate Regulatory Requirements and Design Basis The inspectors identified a finding of very low safety significance for failure to assure that the design basis for the dry cooling tower diesel-driven sump pumps was properly implemented. Specifically, the Train B dry cooling tower diesel-driven sump pump was stored near nonseismic equipment which could fall and damage the pump during an operating-basis earthquake. The dry cooling tower diesel-driven sump pumps are equipment important to safety that are required to protect the ultimate heat sink during a standard project storm coincident with an operating-basis earthquake. The licensee entered this deficiency into their corrective action program for resolution. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to assure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very low safety significance because the finding was a design deficiency that was confirmed not to result in a loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors determined the cause of this finding was not related to a crosscutting element because the performance deficiency does not reflect current operating performance.
Inspection Report# : 2007002 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Recurring Failure of Valve SI-405B to Open A self-revealing violation of very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure to implement effective corrective actions to prevent recurrence of a significant condition adverse to quality. Specifically, on multiple occasions Valve SI-405B failed to stroke open while attempting to place shutdown cooling Train B in service. This violation of Appendix B, Criterion XVI, is being treated as an noncited violation and was entered into the licensees corrective action program. This finding is greater than minor because it affects the the Mitigating Systems Cornerstone attribute of equipment operability, availability, and reliability of systems that respond to initiating events. This finding was evaluated using the significance determination process and was determined to be a finding of very low safety significance because, in each condition identified, it did not represent an actual loss of a safety function. The inspectors also determined that the cause of the condition had crosscutting aspects associated with the corrective action program component in the problem identification and resolution area. This assessment was based on the fact that the licensee failed to thoroughly evaluate the problem such that the resolutions addressed the causes and therefore, corrective actions were inadequate to prevent
 
repetition.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Inspection of Essential Chiller Condenser Tubing A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failing to follow a maintenance procedure during performance of eddy current testing on the safety-related essential chiller Train A condenser tubing. The performance deficiency was the failure to perform a full length eddy current inspection of each tube with an appropriately sized eddy current probe. Subsequently, essential chiller Train A was removed from service to correct a throughwall tube leak in its condenser. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone because the performance deficiency affected the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Appendix A, Phase 1, questions for mitigating systems, the inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, there was no loss of a safety function, and there were no other adverse impacts to the facility. This finding had a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively communicate expectations of procedure compliance.
Inspection Report# : 2006004 (pdf)
Significance: SL-IV Aug 09, 2006 Identified By: NRC Item Type: VIO Violation Inaccurate Performance Indicator Information The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations.
Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRCs recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem identification and resolution crosscutting aspects, in that the implementation of the licensees Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate.
Inspection Report# : 2006009 (pdf)
Significance:        Jul 07, 2006 Identified By: NRC
 
Item Type: VIO Violation Untimely Actions to Reestablish Full Qualification of the Emergency Diesel Generator Starting Air System The inspectors identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." In September 2003, the NRC identified that the emergency diesel generator starting air system was incapable of supplying sufficient air to start its respective emergency diesel generator a minimum of five times without being recharged. To date, the licensee has failed to take appropriate corrective actions in a timely manner to correct this deficiency and restore compliance.
This finding is greater than minor because it affected the mitigating system cornerstone objective due to the degradation of the design-basis capability of the starting air system. This finding has a crosscutting aspect in the corrective action component of the problem identification and resolution area because the licensee failed to take actions to address safety issues in a timely manner, commensurate with its safety significance and complexity. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the starting air system safety function per Generic Letter 91-18 guidance. Additionally, surveillance testing has demonstrated the capability of each diesel generator to start within the required 10 seconds.
Inspection Report# : 2006003 (pdf)
Barrier Integrity Significance:        Feb 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct an Adverse Condition (Welds Not In Accordance With Design)
A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to promptly identify and correct an adverse condition (i.e., steam generator batwing-to-wrapper bar welds not in accordance with design). Specifically, in May 2005, during Refueling Cycle 13, licensee personnel found that the batwing-to-wrapper bar welds were not in accordance with design drawings, but did not enter the adverse condition into the corrective action program until December 2006. This condition was entered into the corrective action program as Condition Report WF3-2006-04395. This finding was more than minor because by not promptly entering the non-conforming welds into the corrective action program and taking actions to correct the adverse condition, it became a more significant condition when two welds failed during Operating Cycle 14. Using the guidance of Appendix J to NRC Inspection Manual Chapter 0609, Significance Determination Process, the finding is determined to have very low safety significance (Green) because there was no tube degradation that exceeded 40 percent through-wall which did not increase in the large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action) program component.
Inspection Report# : 2006012 (pdf)
Emergency Preparedness Significance:        Dec 20, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct a Required Offsite medical Drill in 2005 The inspector identified a noncited violation of 10 CFR 50.54(q) for failure to conduct during 2005 an offsite drill involving a simulated contaminated individual with provision for participation by local medical support services as required by the licensees emergency plan. The licensee failure to conduct the drill is a performance deficiency because the licensee identified the drills postponement in October 2005 and did not appropriately reschedule the drill.
The licensee did not request NRC approval to deviate from this emergency plan requirement. This finding is greater than minor because a degraded proficiency in providing appropriate medical treatment for a contaminated individual
 
has a potential impact on the safety of licensee employees and the public. The finding is of very low safety significance because the licensee failed to conduct only one required drill during the inspection period January 2005 through December 2006, and the drill was not appropriately rescheduled with NRC approval. This finding is a non-cited violation of 10 CFR 50.54(q) and 10 CFR 50 Appendix E, IV, F.1. The licensee has entered this issue into their corrective action system as Condition Report 2006-4429.
Inspection Report# : 2006005 (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 : August 24, 2007
 
Waterford 3 3Q/2007 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Excess Torque Resulting in Pressurizer Skirt Bolt Failures A self-revealing violation of very low safety significance of Technical Specification 6.8.1.a was identified for an inadequate procedure for installing a bolted joint that provided structural support for the pressurizer. Specifically, the installation procedure required applying 8750 ft-lbs torque to make up a bolted joint. Following corrective actions, the licensee discovered that the break away torque on several bolts exceeded 13,400 ft-lbs. The improper bolt tensioning resulted in failure of 1 of 16 bolts and the partial cracking of 3 other bolts that potentially could affect the pressurizers function in a safe shutdown earthquake event. This finding is more than minor because if left uncorrected it could have become a more safety significant concern. The finding was associated with the equipment performance attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations.
This finding was determined to have very low safety significance because a seismic event would not result in a loss-of-coolant accident that exceeded the Technical Specification limit for reactor coolant system leakage. Therefore, this issue screened out in Phase 1 of the MC 0609 significance determination because there was no actual loss of safety function.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure for ESFAS Relay Replacement A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for an inadequate procedure that resulted in the unintentional actuation of five engineered safety features actuation system Train B relays and the loss of a 480 Vac motor control center. The 480 Vac motor control center provided power to the Train B pressurizer heaters and to the control element assembly motor generator Set B. Loss of the control element assembly motor generator increased the likelihood of a reactor trip. This finding is greater than minor because it affects the Initiating Event cornerstone objective procedure quality attribute to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. This finding was evaluated using the significance determination process and was determined to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a crosscutting aspect in the area of human performance associated with resources because the licensee failed to ensure that Work Order 26998 was adequate for the task.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:        Sep 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Boric Acid Leak Evaluation
 
The inspectors identified a noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not following the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not. The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-cutting aspect in the area of human performance, work practices component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(b)).
Inspection Report# : 2007004 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps DRAFT - Green. The team identified a non-cited violation of 10 CFR 50.65(a)(2) for the failure to adequately demonstrate the performance or condition of the dry cooling tower motor-driven sump pumps. Specifically, the licensee failed to periodically verify that the pump flow rates were consistent with their design basis requirements and pump performance problems were likely to go unnoticed. Therefore, the licensee had no technical justification for continued Maintenance Rule (a)(2) status.
Failure to develop and implement technically justifiable performance criteria for the motor-driven sump pumps, for compliance with provisions of the Maintenance Rule, was a performance deficiency. The finding was greater than minor because it could be a more significant safety concern if left uncorrected. In addition, the finding was similar to non-minor finding Example 7.b in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there were performance concerns associated with the dry cooling tower sump pumps. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement FME Procedure for Dry Cooling Tower Sumps DRAFT - The team identified a finding for the failure to properly implement the site foreign material exclusion procedure for the dry cooling tower sumps. Specifically, the procedure required the establishment of a foreign material exclusion area if foreign materials could adversely impact equipment function. The area surrounding the dry cooling tower sumps met this criteria but the licensee failed to establish a foreign material exclusion area to protect the sump pump system from damage. The sump pumps had previously suffered damage due to foreign material intrusion.
The failure to properly implement the site foreign material exclusion procedure was a performance deficiency. The finding was more than minor because it affected the mitigating systems cornerstone objective (external factors attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of human performance (work practices component) in that personnel failed to
 
follow a site procedure (H.4(b)). The finding was indicative of current plant performance because the open sump and the foreign material vulnerability was known to plant personnel on an ongoing basis Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Restoring Power to Dry Cooling Tower Sump Pumps DRAFT - The team identified a non-cited violation of Technical Specification 6.8.1.a, Procedures, for inadequate procedural guidance for operators to respond to a postulated loss of offsite power event coincident with a design basis rain event. The design basis calculation specified that, during certain rain precipitation events, operators must transfer the pump power to a safety related power source within 30 minutes of a loss of offsite power to protect a safety related motor control center from flooding. The motor control centers are needed to ensure ultimate heat sink operability.
During plant walkdowns, due to the sequencing of steps in the procedure, operators took approximately 50 minutes to transfer essential power to the pumps. In addition, the procedural step was worded inappropriately because it allowed operators to wait the full 30 minutes before starting the action.
The failure to provide an emergency operating procedure that could be consistently completed within the required time limits was a performance deficiency. This finding was more than minor because it affected the mitigating systems cornerstone objective (external factors component) to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding was similar to non-minor finding Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was reasonable doubt of the operability of the system under certain heavy rain conditions. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the team determined that a Phase 2 significance determination was required because the finding potentially represented a loss of system safety function.
The team performed a Phase 2 significance determination and found the finding was potentially greater than Green in significance. A Region IV senior reactor analyst performed a Phase 3 significance determination and found the issue was of very low safety significance.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria for Battery Cell-to-Cell and Terminal Connection Resistance Value DRAFT - The Team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure that the 125 Vdc safety-related batteries would remain operable if all the intercell and terminal connections were at the resistance value of 150 micro-ohms as allowed by Technical Specification Surveillance Requirement 4.8.2.1.b.2 and 4.8.2.1.c.3.
The failure to adequately verify or check a design value in accordance with NRC design control requirements was a performance deficiency. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Prompt Corrective Measures to Address Degraded Dry Cooling Towers DRAFT - The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly correct a condition adverse to quality (dirt and debris in the dry cooling tower heat exchanger fins). The condition adversely impacted the heat exchangers heat removal rates. The dry cooling towers had very
 
little design margin under some scenarios. In addition, the licensee failed to respond to trend data that showed degraded heat exchanger performance, had no basis for the specified 5 year cleaning interval specified in their heat exchanger program, and hadnt actually cleaned the towers for approximately 11 years. This issue was entered into the licensee's corrective action program as Condition Report CR-WF3-2007-01433.
This finding was more than minor because it was similar to non-minor Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt of the operability of the dry cooling towers. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because the finding was a qualification deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of Problem Identification and Resolution (corrective action program attribute) in that the issue was identified but corrective actions were not taken in a prompt manner (P.1(d)). The issue was indicative of current performance because the system engineer was aware of the degraded cooling tower condition for several years.
Inspection Report# : 2007007 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Design Basis into Drawings The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III, Design Control, for failure to assure that the design basis, as specified in the license application, was correctly translated into drawings and the actual plant configuration. Specifically, Waterford Final Safety Analysis Report, Section 2.4.2.3.3.d, describes openings in the dry cooling tower cubicles that help preclude the possibility of flooding Motor Control Centers 3A315-S and 3B315-S during the probable maximum precipitation event. These openings serve as a backup to the floor drains located in each cubicle. Current plant configuration and Drawing G-499 S06, Common Foundation Structure, Masonry, Sheet 6, do not conform to the design basis, in that there are no openings other than the floor drains. These motor control centers control power to the wet and dry cooling tower fans, which act as the ultimate heat sink. The licensee entered this issue into their corrective action program for resolution. This finding is more than minor because it is associated with the design control attribute and affects the Mitigating Systems cornerstone objective to ensure the reliability of the dry cooling tower system during the probable maximum precipitation event on the plant site. The normal floor drains had historically clogged and the drainage openings were needed to limit flood related challenges to the motor control centers. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the wet and dry cooling tower systems safety functions during the past year per Part 9900: Technical Guidance, Operability Determinations & Functional Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality.
Inspection Report# : 2007002 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: FIN Finding Failure to Ensure that Written Procedures Adequately Incorporate Regulatory Requirements and Design Basis The inspectors identified a finding of very low safety significance for failure to assure that the design basis for the dry cooling tower diesel-driven sump pumps was properly implemented. Specifically, the Train B dry cooling tower diesel-driven sump pump was stored near nonseismic equipment which could fall and damage the pump during an operating-basis earthquake. The dry cooling tower diesel-driven sump pumps are equipment important to safety that are required to protect the ultimate heat sink during a standard project storm coincident with an operating-basis earthquake. The licensee entered this deficiency into their corrective action program for resolution. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to assure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very low safety significance because the finding was a design deficiency that was confirmed not to result in a loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors determined the cause of this finding was not related to a crosscutting element because the performance deficiency does not reflect current
 
operating performance.
Inspection Report# : 2007002 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Recurring Failure of Valve SI-405B to Open A self-revealing violation of very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure to implement effective corrective actions to prevent recurrence of a significant condition adverse to quality. Specifically, on multiple occasions Valve SI-405B failed to stroke open while attempting to place shutdown cooling Train B in service. This violation of Appendix B, Criterion XVI, is being treated as an noncited violation and was entered into the licensees corrective action program. This finding is greater than minor because it affects the the Mitigating Systems Cornerstone attribute of equipment operability, availability, and reliability of systems that respond to initiating events. This finding was evaluated using the significance determination process and was determined to be a finding of very low safety significance because, in each condition identified, it did not represent an actual loss of a safety function. The inspectors also determined that the cause of the condition had crosscutting aspects associated with the corrective action program component in the problem identification and resolution area. This assessment was based on the fact that the licensee failed to thoroughly evaluate the problem such that the resolutions addressed the causes and therefore, corrective actions were inadequate to prevent repetition.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 07, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Inspection of Essential Chiller Condenser Tubing A self-revealing noncited violation of Technical Specification 6.8.1.a was identified for failing to follow a maintenance procedure during performance of eddy current testing on the safety-related essential chiller Train A condenser tubing. The performance deficiency was the failure to perform a full length eddy current inspection of each tube with an appropriately sized eddy current probe. Subsequently, essential chiller Train A was removed from service to correct a throughwall tube leak in its condenser. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone because the performance deficiency affected the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Appendix A, Phase 1, questions for mitigating systems, the inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, there was no loss of a safety function, and there were no other adverse impacts to the facility. This finding had a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively communicate expectations of procedure compliance.
Inspection Report# : 2006004 (pdf)
Significance: SL-IV Aug 09, 2006 Identified By: NRC Item Type: VIO Violation Inaccurate Performance Indicator Information The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations.
Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight
 
with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRCs recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem identification and resolution crosscutting aspects, in that the implementation of the licensees Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate.
Inspection Report# : 2006009 (pdf)
Barrier Integrity Significance:      Feb 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct an Adverse Condition (Welds Not In Accordance With Design)
A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to promptly identify and correct an adverse condition (i.e., steam generator batwing-to-wrapper bar welds not in accordance with design). Specifically, in May 2005, during Refueling Cycle 13, licensee personnel found that the batwing-to-wrapper bar welds were not in accordance with design drawings, but did not enter the adverse condition into the corrective action program until December 2006. This condition was entered into the corrective action program as Condition Report WF3-2006-04395. This finding was more than minor because by not promptly entering the non-conforming welds into the corrective action program and taking actions to correct the adverse condition, it became a more significant condition when two welds failed during Operating Cycle 14. Using the guidance of Appendix J to NRC Inspection Manual Chapter 0609, Significance Determination Process, the finding is determined to have very low safety significance (Green) because there was no tube degradation that exceeded 40 percent through-wall which did not increase in the large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action) program component.
Inspection Report# : 2006012 (pdf)
Emergency Preparedness Significance:      Dec 20, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conduct a Required Offsite medical Drill in 2005 The inspector identified a noncited violation of 10 CFR 50.54(q) for failure to conduct during 2005 an offsite drill involving a simulated contaminated individual with provision for participation by local medical support services as required by the licensees emergency plan. The licensee failure to conduct the drill is a performance deficiency because the licensee identified the drills postponement in October 2005 and did not appropriately reschedule the drill.
The licensee did not request NRC approval to deviate from this emergency plan requirement. This finding is greater
 
than minor because a degraded proficiency in providing appropriate medical treatment for a contaminated individual has a potential impact on the safety of licensee employees and the public. The finding is of very low safety significance because the licensee failed to conduct only one required drill during the inspection period January 2005 through December 2006, and the drill was not appropriately rescheduled with NRC approval. This finding is a non-cited violation of 10 CFR 50.54(q) and 10 CFR 50 Appendix E, IV, F.1. The licensee has entered this issue into their corrective action system as Condition Report 2006-4429.
Inspection Report# : 2006005 (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 : December 07, 2007
 
Waterford 3 4Q/2007 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for a Fire in Vital Switchgear Room B The inspectors identified two examples of a noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility. In the first example, the pre-fire strategy for vital switchgear Room B did not contain adequate information regarding the doors required to be open to allow the desired ventilation flowpath, nor did it contain the required number of smoke ejectors necessary to desmoke the switchgear room in a manner that would allow the implementation of OP-901-524, Fire In Areas Affecting Safe Shutdown. In the second example, the licensee did not take corrective actions for a previously identified issue in a timely fashion. Specifically, the deficiencies in the pre-fire strategy for vital switchgear Room B were first identified on August 21, 2006. The deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the non-conformance with licensee management. The licensee entered this deficiency into their corrective action program for resolution. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution.
Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)).
Inspection Report# : 2007004 (pdf)
Significance:        Sep 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Boric Acid Leak Evaluation The inspectors identified a noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not following the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not. The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-cutting aspect in the area of human performance, work practices component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(b)).
 
Inspection Report# : 2007004 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps DRAFT - Green. The team identified a non-cited violation of 10 CFR 50.65(a)(2) for the failure to adequately demonstrate the performance or condition of the dry cooling tower motor-driven sump pumps. Specifically, the licensee failed to periodically verify that the pump flow rates were consistent with their design basis requirements and pump performance problems were likely to go unnoticed. Therefore, the licensee had no technical justification for continued Maintenance Rule (a)(2) status.
Failure to develop and implement technically justifiable performance criteria for the motor-driven sump pumps, for compliance with provisions of the Maintenance Rule, was a performance deficiency. The finding was greater than minor because it could be a more significant safety concern if left uncorrected. In addition, the finding was similar to non-minor finding Example 7.b in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there were performance concerns associated with the dry cooling tower sump pumps. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement FME Procedure for Dry Cooling Tower Sumps DRAFT - The team identified a finding for the failure to properly implement the site foreign material exclusion procedure for the dry cooling tower sumps. Specifically, the procedure required the establishment of a foreign material exclusion area if foreign materials could adversely impact equipment function. The area surrounding the dry cooling tower sumps met this criteria but the licensee failed to establish a foreign material exclusion area to protect the sump pump system from damage. The sump pumps had previously suffered damage due to foreign material intrusion.
The failure to properly implement the site foreign material exclusion procedure was a performance deficiency. The finding was more than minor because it affected the mitigating systems cornerstone objective (external factors attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of human performance (work practices component) in that personnel failed to follow a site procedure (H.4(b)). The finding was indicative of current plant performance because the open sump and the foreign material vulnerability was known to plant personnel on an ongoing basis Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Restoring Power to Dry Cooling Tower Sump Pumps DRAFT - The team identified a non-cited violation of Technical Specification 6.8.1.a, Procedures, for inadequate procedural guidance for operators to respond to a postulated loss of offsite power event coincident with a design basis rain event. The design basis calculation specified that, during certain rain precipitation events, operators must transfer the pump power to a safety related power source within 30 minutes of a loss of offsite power to protect a safety related motor control center from flooding. The motor control centers are needed to ensure ultimate heat sink operability.
 
During plant walkdowns, due to the sequencing of steps in the procedure, operators took approximately 50 minutes to transfer essential power to the pumps. In addition, the procedural step was worded inappropriately because it allowed operators to wait the full 30 minutes before starting the action.
The failure to provide an emergency operating procedure that could be consistently completed within the required time limits was a performance deficiency. This finding was more than minor because it affected the mitigating systems cornerstone objective (external factors component) to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding was similar to non-minor finding Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was reasonable doubt of the operability of the system under certain heavy rain conditions. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the team determined that a Phase 2 significance determination was required because the finding potentially represented a loss of system safety function.
The team performed a Phase 2 significance determination and found the finding was potentially greater than Green in significance. A Region IV senior reactor analyst performed a Phase 3 significance determination and found the issue was of very low safety significance.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria for Battery Cell-to-Cell and Terminal Connection Resistance Value DRAFT - The Team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure that the 125 Vdc safety-related batteries would remain operable if all the intercell and terminal connections were at the resistance value of 150 micro-ohms as allowed by Technical Specification Surveillance Requirement 4.8.2.1.b.2 and 4.8.2.1.c.3.
The failure to adequately verify or check a design value in accordance with NRC design control requirements was a performance deficiency. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Prompt Corrective Measures to Address Degraded Dry Cooling Towers DRAFT - The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly correct a condition adverse to quality (dirt and debris in the dry cooling tower heat exchanger fins). The condition adversely impacted the heat exchangers heat removal rates. The dry cooling towers had very little design margin under some scenarios. In addition, the licensee failed to respond to trend data that showed degraded heat exchanger performance, had no basis for the specified 5 year cleaning interval specified in their heat exchanger program, and hadnt actually cleaned the towers for approximately 11 years. This issue was entered into the licensee's corrective action program as Condition Report CR-WF3-2007-01433.
This finding was more than minor because it was similar to non-minor Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt of the operability of the dry cooling towers. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because the finding was a qualification deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of Problem Identification and Resolution (corrective action program attribute) in that the issue was identified but corrective actions were not taken in a prompt manner (P.1(d)). The issue was indicative of current performance because the system engineer was aware of the degraded cooling tower condition for several years.
 
Inspection Report# : 2007007 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Design Basis into Drawings The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III, Design Control, for failure to assure that the design basis, as specified in the license application, was correctly translated into drawings and the actual plant configuration. Specifically, Waterford Final Safety Analysis Report, Section 2.4.2.3.3.d, describes openings in the dry cooling tower cubicles that help preclude the possibility of flooding Motor Control Centers 3A315-S and 3B315-S during the probable maximum precipitation event. These openings serve as a backup to the floor drains located in each cubicle. Current plant configuration and Drawing G-499 S06, Common Foundation Structure, Masonry, Sheet 6, do not conform to the design basis, in that there are no openings other than the floor drains. These motor control centers control power to the wet and dry cooling tower fans, which act as the ultimate heat sink. The licensee entered this issue into their corrective action program for resolution. This finding is more than minor because it is associated with the design control attribute and affects the Mitigating Systems cornerstone objective to ensure the reliability of the dry cooling tower system during the probable maximum precipitation event on the plant site. The normal floor drains had historically clogged and the drainage openings were needed to limit flood related challenges to the motor control centers. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the wet and dry cooling tower systems safety functions during the past year per Part 9900: Technical Guidance, Operability Determinations & Functional Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality.
Inspection Report# : 2007002 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: FIN Finding Failure to Ensure that Written Procedures Adequately Incorporate Regulatory Requirements and Design Basis The inspectors identified a finding of very low safety significance for failure to assure that the design basis for the dry cooling tower diesel-driven sump pumps was properly implemented. Specifically, the Train B dry cooling tower diesel-driven sump pump was stored near nonseismic equipment which could fall and damage the pump during an operating-basis earthquake. The dry cooling tower diesel-driven sump pumps are equipment important to safety that are required to protect the ultimate heat sink during a standard project storm coincident with an operating-basis earthquake. The licensee entered this deficiency into their corrective action program for resolution. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to assure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very low safety significance because the finding was a design deficiency that was confirmed not to result in a loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors determined the cause of this finding was not related to a crosscutting element because the performance deficiency does not reflect current operating performance.
Inspection Report# : 2007002 (pdf)
Barrier Integrity Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Missed Reactor Coolant System Chemistry Samples
 
The inspectors identified a noncited violation of Technical Specification (TS) 3.4.7 for multiple failures to complete a radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity.
Specifically, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the TS required interval of 136 to 229 days. EBAR is the average of the sum of average beta and gamma energies per disintegration for isotopes, other than radioiodines, with half-lives greater than fifteen minutes. Daily RCS samples are compared to this calculated value in order to ensure that 10CFR50.67 dose limits at the site boundary are not exceeded in the event of an accident scenario. The licensee entered this issue into their corrective action program for resolution. The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier. This finding had a crosscutting aspect in the area of human performance. Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)).
Inspection Report# : 2007004 (pdf)
Significance:      Feb 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct an Adverse Condition (Welds Not In Accordance With Design)
A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to promptly identify and correct an adverse condition (i.e., steam generator batwing-to-wrapper bar welds not in accordance with design). Specifically, in May 2005, during Refueling Cycle 13, licensee personnel found that the batwing-to-wrapper bar welds were not in accordance with design drawings, but did not enter the adverse condition into the corrective action program until December 2006. This condition was entered into the corrective action program as Condition Report WF3-2006-04395. This finding was more than minor because by not promptly entering the non-conforming welds into the corrective action program and taking actions to correct the adverse condition, it became a more significant condition when two welds failed during Operating Cycle 14. Using the guidance of Appendix J to NRC Inspection Manual Chapter 0609, Significance Determination Process, the finding is determined to have very low safety significance (Green) because there was no tube degradation that exceeded 40 percent through-wall which did not increase in the large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action) program component.[P.1(a)]
Inspection Report# : 2006012 (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 : February 04, 2008
 
Waterford 3 1Q/2008 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion V was identified for an inadequate maintenance procedure. Specifically, MM-006-054, Check Valve Inspection (Tilting Disc), lacked sufficient detail to prevent poor workmanship during maintenance on safety injection Tank 1A discharge check Valve SI-335A. This poor workmanship allowed Valve SI-335A to be reassembled with a cocked hinge pin cover, resulting in reactor coolant system (RCS) leakage. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the procedure quality attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because assuming worst case degradation, the Valve SI-335A leak would not result in exceeding the Technical Specification limit for identified RCS leakage. This finding had a crosscutting aspect in the resources component of the human performance area. Specifically, the licensee failed to provide the maintenance technician with a complete and accurate maintenance procedure [H.2(c)].
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Reactor Coolant Pump 1A Seal Leak Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion XVI was identified for the licensees failure to promptly identify and correct a significant condition adverse to quality. Specifically, the licensee did not identify a seal leak on reactor coolant Pump 1A in a timely fashion. During efforts to identify the source of leakage, the licensee effectively ruled out the reactor coolant pump seal areas based on an incorrect assumption. When no other significant sources of leakage could be found, the decision was made to monitor the leakage and take no further actions until the mid-cycle outage. This unidentified reactor coolant system leakage caused degradation to the reactor coolant pump cover, main casing stud nuts, shroud wall, and carbon steel flanges. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the equipment performance attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available. This finding had a crosscutting aspect in the area of human performance associated with decision-making in that the licensee did not use conservative assumptions in the reactor coolant system leakage investigation [H.1(b)].
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Dec 31, 2007 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Follow Procedure Review Process Green. The inspectors identified a Green noncited violation of Technical Specification 6.8.1.a (Procedures) for failure to correctly implement a procedure recommended in Appendix A of Regulatory Guide 1.33. Specifically, the failure to follow Site Procedure W2.109, Procedure Development, Review, and Approval, led to the unapproved deletion of the Special Scope section of the Quality Assurance Program Manual. The Special Scope section contained the fire protection quality assurance (QA) program components and discussion for their implementation. This deleted information is required by the Waterford 3 Steam Electric Station License Condition 2.C.9. The licensee entered this issue into their corrective action program for resolution.
The finding was more than minor because if left uncorrected, it would become a more significant safety concern.
Using Inspection Manual Chapter 0609, Appendix F, this finding can be assigned a low degradation rating and screen as green, since current QA audit standards contain a similar level of detail as the criteria contained in the deleted Special Scope document.
Inspection Report# : 2007005 (pdf)
Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for a Fire in Vital Switchgear Room B The inspectors identified two examples of a noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility. In the first example, the pre-fire strategy for vital switchgear Room B did not contain adequate information regarding the doors required to be open to allow the desired ventilation flowpath, nor did it contain the required number of smoke ejectors necessary to desmoke the switchgear room in a manner that would allow the implementation of OP-901-524, Fire In Areas Affecting Safe Shutdown. In the second example, the licensee did not take corrective actions for a previously identified issue in a timely fashion. Specifically, the deficiencies in the pre-fire strategy for vital switchgear Room B were first identified on August 21, 2006. The deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the non-conformance with licensee management. The licensee entered this deficiency into their corrective action program for resolution. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution.
Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)).
Inspection Report# : 2007004 (pdf)
Significance:        Sep 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Boric Acid Leak Evaluation The inspectors identified a noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not following the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not. The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-
 
cutting aspect in the area of human performance, work practices component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(b)).
Inspection Report# : 2007004 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps Failure to Meet Maintenance Rule Requirements for Dry Cooling Tower Sump Pumps DRAFT - Green. The team identified a non-cited violation of 10 CFR 50.65(a)(2) for the failure to adequately demonstrate the performance or condition of the dry cooling tower motor-driven sump pumps. Specifically, the licensee failed to periodically verify that the pump flow rates were consistent with their design basis requirements and pump performance problems were likely to go unnoticed. Therefore, the licensee had no technical justification for continued Maintenance Rule (a)(2) status.
Failure to develop and implement technically justifiable performance criteria for the motor-driven sump pumps, for compliance with provisions of the Maintenance Rule, was a performance deficiency. The finding was greater than minor because it could be a more significant safety concern if left uncorrected. In addition, the finding was similar to non-minor finding Example 7.b in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there were performance concerns associated with the dry cooling tower sump pumps. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement FME Procedure for Dry Cooling Tower Sumps DRAFT - The team identified a finding for the failure to properly implement the site foreign material exclusion procedure for the dry cooling tower sumps. Specifically, the procedure required the establishment of a foreign material exclusion area if foreign materials could adversely impact equipment function. The area surrounding the dry cooling tower sumps met this criteria but the licensee failed to establish a foreign material exclusion area to protect the sump pump system from damage. The sump pumps had previously suffered damage due to foreign material intrusion.
The failure to properly implement the site foreign material exclusion procedure was a performance deficiency. The finding was more than minor because it affected the mitigating systems cornerstone objective (external factors attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of human performance (work practices component) in that personnel failed to follow a site procedure (H.4(b)). The finding was indicative of current plant performance because the open sump and the foreign material vulnerability was known to plant personnel on an ongoing basis Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Restoring Power to Dry Cooling Tower Sump Pumps DRAFT - The team identified a non-cited violation of Technical Specification 6.8.1.a, Procedures, for inadequate procedural guidance for operators to respond to a postulated loss of offsite power event coincident with a design basis rain event. The design basis calculation specified that, during certain rain precipitation events, operators must transfer the pump power to a safety related power source within 30 minutes of a loss of offsite power to protect a safety related
 
motor control center from flooding. The motor control centers are needed to ensure ultimate heat sink operability.
During plant walkdowns, due to the sequencing of steps in the procedure, operators took approximately 50 minutes to transfer essential power to the pumps. In addition, the procedural step was worded inappropriately because it allowed operators to wait the full 30 minutes before starting the action.
The failure to provide an emergency operating procedure that could be consistently completed within the required time limits was a performance deficiency. This finding was more than minor because it affected the mitigating systems cornerstone objective (external factors component) to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding was similar to non-minor finding Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was reasonable doubt of the operability of the system under certain heavy rain conditions. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the team determined that a Phase 2 significance determination was required because the finding potentially represented a loss of system safety function.
The team performed a Phase 2 significance determination and found the finding was potentially greater than Green in significance. A Region IV senior reactor analyst performed a Phase 3 significance determination and found the issue was of very low safety significance.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria for Battery Cell-to-Cell and Terminal Connection Resistance Value DRAFT - The Team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure that the 125 Vdc safety-related batteries would remain operable if all the intercell and terminal connections were at the resistance value of 150 micro-ohms as allowed by Technical Specification Surveillance Requirement 4.8.2.1.b.2 and 4.8.2.1.c.3.
The failure to adequately verify or check a design value in accordance with NRC design control requirements was a performance deficiency. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2007007 (pdf)
Significance:      May 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Prompt Corrective Measures to Address Degraded Dry Cooling Towers DRAFT - The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly correct a condition adverse to quality (dirt and debris in the dry cooling tower heat exchanger fins). The condition adversely impacted the heat exchangers heat removal rates. The dry cooling towers had very little design margin under some scenarios. In addition, the licensee failed to respond to trend data that showed degraded heat exchanger performance, had no basis for the specified 5 year cleaning interval specified in their heat exchanger program, and hadnt actually cleaned the towers for approximately 11 years. This issue was entered into the licensee's corrective action program as Condition Report CR-WF3-2007-01433.
This finding was more than minor because it was similar to non-minor Example 3.k in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt of the operability of the dry cooling towers. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because the finding was a qualification deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had a crosscutting aspect in the area of Problem Identification and Resolution (corrective action program attribute) in that the issue was identified but corrective actions were not taken in a prompt manner (P.1(d)). The issue was indicative of current performance because the system engineer was aware of the degraded cooling tower condition for several years.
 
Inspection Report# : 2007007 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Design Basis into Drawings The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III, Design Control, for failure to assure that the design basis, as specified in the license application, was correctly translated into drawings and the actual plant configuration. Specifically, Waterford Final Safety Analysis Report, Section 2.4.2.3.3.d, describes openings in the dry cooling tower cubicles that help preclude the possibility of flooding Motor Control Centers 3A315-S and 3B315-S during the probable maximum precipitation event. These openings serve as a backup to the floor drains located in each cubicle. Current plant configuration and Drawing G-499 S06, Common Foundation Structure, Masonry, Sheet 6, do not conform to the design basis, in that there are no openings other than the floor drains. These motor control centers control power to the wet and dry cooling tower fans, which act as the ultimate heat sink. The licensee entered this issue into their corrective action program for resolution. This finding is more than minor because it is associated with the design control attribute and affects the Mitigating Systems cornerstone objective to ensure the reliability of the dry cooling tower system during the probable maximum precipitation event on the plant site. The normal floor drains had historically clogged and the drainage openings were needed to limit flood related challenges to the motor control centers. The finding was determined to be of very low safety significance because the deficiency did not represent an actual loss of the wet and dry cooling tower systems safety functions during the past year per Part 9900: Technical Guidance, Operability Determinations & Functional Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality.
Inspection Report# : 2007002 (pdf)
Significance:        Apr 07, 2007 Identified By: NRC Item Type: FIN Finding Failure to Ensure that Written Procedures Adequately Incorporate Regulatory Requirements and Design Basis The inspectors identified a finding of very low safety significance for failure to assure that the design basis for the dry cooling tower diesel-driven sump pumps was properly implemented. Specifically, the Train B dry cooling tower diesel-driven sump pump was stored near nonseismic equipment which could fall and damage the pump during an operating-basis earthquake. The dry cooling tower diesel-driven sump pumps are equipment important to safety that are required to protect the ultimate heat sink during a standard project storm coincident with an operating-basis earthquake. The licensee entered this deficiency into their corrective action program for resolution. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to assure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very low safety significance because the finding was a design deficiency that was confirmed not to result in a loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The inspectors determined the cause of this finding was not related to a crosscutting element because the performance deficiency does not reflect current operating performance.
Inspection Report# : 2007002 (pdf)
Barrier Integrity Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Missed Reactor Coolant System Chemistry Samples The inspectors identified a noncited violation of Technical Specification (TS) 3.4.7 for multiple failures to complete a
 
radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity.
Specifically, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the TS required interval of 136 to 229 days. EBAR is the average of the sum of average beta and gamma energies per disintegration for isotopes, other than radioiodines, with half-lives greater than fifteen minutes. Daily RCS samples are compared to this calculated value in order to ensure that 10CFR50.67 dose limits at the site boundary are not exceeded in the event of an accident scenario. The licensee entered this issue into their corrective action program for resolution. The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier. This finding had a crosscutting aspect in the area of human performance. Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)).
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Current Radiological Information Prior to Entering a High Radiation Area Green. The inspector reviewed a self-revealing, noncited violation of Technical Specification 6.12.1.b that resulted when workers did not obtain current radiological information before entering a high radiation area as required by the Technical Specifications. On December 12, 2006, two workers accessed a high radiation area near the Reactor Coolant Pump 1B Cold Leg through a pathway not discussed with radiation protection and received electronic dose rate alarms. Upon investigation, the licensee determined the workers did not clearly communicate the work scope and the travel path for accessing the work areas; therefore, the workers were not briefed for the radiological conditions of the areas near the Reactor Coolant Pump 1B Cold Leg. The peak dose rates for the two workers were 210 millirem per hour and 361 millirem per hour, respectively. Corrective actions implemented by the licensee were that the workers completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and received additional coaching from radiation protection personnel.
The failure to obtain current radiological information prior to entering a high radiation area is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not obtaining high radiation area dose rates does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self and peer checking, when discussing the work scope and work areas with radiation protection staff [H.4.(a)].
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Green. The inspector reviewed two examples of a self-revealing, noncited violation of Technical Specification 5.4.1
 
that resulted when workers failed to follow their radiation work permit instructions. The first example occurred on October 11, 2007, when an operator accessed Valves RC 109 and RC 110 by a travel path not discussed with radiation protection personnel and without obtaining current radiological conditions as specified in the radiation work permit.
As the operator passed through the pipe-chase to access the valves, the worker received a dose rate alarm. The highest dose rate levels were 80 millirem per hour along the travel path. The second example occurred on October 12, 2007, when a maintenance mechanic entered the Safeguards B room without a current radiological briefing as specified in the radiation work permit. Radiation protection personnel requested the worker wait to access Safeguards A room while the radiological conditions were changing (shutdown cooling in progress) and did not know the worker also needed to access the B room. The worker, who had previously entered the B room but failed to realize this room also had changing radiological conditions, did not receive current radiological conditions for this room and received a dose rate alarm. The workers peak dose rate was 61 millirem per hour. The licensees corrective actions for the first example were that a radiation protection supervisor conducted an interview with worker, and the worker completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and human performance error review. For the second example, the immediate corrective action was to exclude the individual from the radiological controlled area then perform a human performance error review.
The failure to follow a radiation work permit instruction is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not following their radiation work permit does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self checking, to ensure the full work scope, locations, and radiological conditions were discussed with radiation protection personnel as required by the radiation work permit [H.4.(a)].
Inspection Report# : 2007005 (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 Last modified : June 05, 2008
 
Waterford 3 2Q/2008 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion V was identified for an inadequate maintenance procedure. Specifically, MM-006-054, Check Valve Inspection (Tilting Disc), lacked sufficient detail to prevent poor workmanship during maintenance on safety injection Tank 1A discharge check Valve SI-335A. This poor workmanship allowed Valve SI-335A to be reassembled with a cocked hinge pin cover, resulting in reactor coolant system (RCS) leakage. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the procedure quality attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because assuming worst case degradation, the Valve SI-335A leak would not result in exceeding the Technical Specification limit for identified RCS leakage. This finding had a crosscutting aspect in the resources component of the human performance area. Specifically, the licensee failed to provide the maintenance technician with a complete and accurate maintenance procedure [H.2(c)].
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Reactor Coolant Pump 1A Seal Leak Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion XVI was identified for the licensees failure to promptly identify and correct a significant condition adverse to quality. Specifically, the licensee did not identify a seal leak on reactor coolant Pump 1A in a timely fashion. During efforts to identify the source of leakage, the licensee effectively ruled out the reactor coolant pump seal areas based on an incorrect assumption. When no other significant sources of leakage could be found, the decision was made to monitor the leakage and take no further actions until the mid-cycle outage. This unidentified reactor coolant system leakage caused degradation to the reactor coolant pump cover, main casing stud nuts, shroud wall, and carbon steel flanges. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the equipment performance attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available. This finding had a crosscutting aspect in the area of human performance associated with decision-making in that the licensee did not use conservative assumptions in the reactor coolant system leakage investigation [H.1(b)].
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Apr 07, 2008 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to re-evaluate previously identified boric acid leaks The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions for a condition adverse to quality. Specifically, the licensee developed a corrective action to evaluate the condition of existing boric acid leaks. However, the effort failed to identify and evaluate multiple existing boric acid leaks on safety related components, including some that had deteriorated since initial discovery. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2007 3951.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. Specifically, some unchecked boric acid leaks may have worsened and corroded safety related equipment.
Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding had very low risk significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because engineers failed to implement proper error prevention techniques when identifying boric acid leaks for additional review H.4 (a).
Inspection Report# : 2008002 (pdf)
Significance:      Apr 07, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to correct "Fuel Oil Receipt and Transfer" procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to implement adequate corrective actions for a deficient emergency diesel generator fuel tank filling procedure (a condition adverse to quality). The licensee had identified the deficiency following a previous event when fuel oil leaked out of multiple fuel oil injectors during a diesel run. Procedural steps were needed to adequately vent the fill line following pressurization during fuel oil tank filling. However, the licensee only corrected the procedure in one section and, when a different section was used, the problem reoccurred. The fuel oil leak led to the emergency diesel generator being declared inoperable. In addition, the fuel oil created a potential fire hazard. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008 1345.
The finding was more than minor because it was similar to nonminor example 4.f in Inspection Manual Chapter 0612, "Examples of Minor Issues," in that emergency diesel generator operability was affected. Further, the oil created a fire hazard. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:      Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Essential chiller AB return Header B Isolation Valve CHW 786B Misposition The inspectors documented a self revealing noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to correctly position a valve during a surveillance. The procedure required operators to position the essential Chiller AB return Header B isolation Valve CHW 786B closed but operators left the valve in the open position. This resulted in cross connecting the essential services chilled water Loops A and B, which led to an unplanned entry into Technical Specifications 3.7.12 and 3.0.3. The violation was revealed through a control room alarm. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0778.
The finding was more than minor because, if left uncorrected, would have become a more significant safety concern.
Specifically, with both loops of the essential services chilled water system cross connected, the system was no longer single-failure proof. A leak in one of the essential chilled water loops would have caused both units to become inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the
 
issue screened as having very low safety significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because operators failed to implement self-checking techniques when performing procedure steps H.4(a).
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation ACCW pump failure due to inaccurate operator aid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to correct a condition adverse to quality (inadequate instructions that led to low fuel oil and the failure of auxiliary component cooling water pump bearing). Specifically, the licensee's corrective action for a previous event called for an operator aid (oil level label). However, the operator aid contained incorrect and confusing information. Consequently, another auxiliary component cooling water pump failed. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0350.
The finding was more than minor because it was similar to nonminor violation example 4.f in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that the problem affected auxiliary component cooling water Pump B operability. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had crosscutting aspects associated with Human Performance area, resources program component, because the licensee failed to have correct labeling on components H.2(c).
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Review Process Green. The inspectors identified a Green noncited violation of Technical Specification 6.8.1.a (Procedures) for failure to correctly implement a procedure recommended in Appendix A of Regulatory Guide 1.33. Specifically, the failure to follow Site Procedure W2.109, Procedure Development, Review, and Approval, led to the unapproved deletion of the Special Scope section of the Quality Assurance Program Manual. The Special Scope section contained the fire protection quality assurance (QA) program components and discussion for their implementation. This deleted information is required by the Waterford 3 Steam Electric Station License Condition 2.C.9. The licensee entered this issue into their corrective action program for resolution.
The finding was more than minor because if left uncorrected, it would become a more significant safety concern.
Using Inspection Manual Chapter 0609, Appendix F, this finding can be assigned a low degradation rating and screen as green, since current QA audit standards contain a similar level of detail as the criteria contained in the deleted Special Scope document.
Inspection Report# : 2007005 (pdf)
Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for a Fire in Vital Switchgear Room B The inspectors identified two examples of a noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility. In the first example, the pre-fire strategy for vital switchgear Room B did not contain adequate information regarding the doors required to be open to
 
allow the desired ventilation flowpath, nor did it contain the required number of smoke ejectors necessary to desmoke the switchgear room in a manner that would allow the implementation of OP-901-524, Fire In Areas Affecting Safe Shutdown. In the second example, the licensee did not take corrective actions for a previously identified issue in a timely fashion. Specifically, the deficiencies in the pre-fire strategy for vital switchgear Room B were first identified on August 21, 2006. The deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the non-conformance with licensee management. The licensee entered this deficiency into their corrective action program for resolution. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution.
Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)).
Inspection Report# : 2007004 (pdf)
Significance:        Sep 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Boric Acid Leak Evaluation The inspectors identified a noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not following the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not. The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-cutting aspect in the area of human performance, work practices component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(b)).
Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Missed Reactor Coolant System Chemistry Samples The inspectors identified a noncited violation of Technical Specification (TS) 3.4.7 for multiple failures to complete a radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity.
Specifically, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the TS required interval of 136 to 229 days. EBAR is the average of the sum of average beta and gamma energies per disintegration for isotopes, other than radioiodines, with half-lives greater than fifteen minutes. Daily RCS samples are compared to this calculated value in order to ensure that 10CFR50.67 dose limits at the site boundary are not exceeded in the event of an accident scenario. The licensee entered this issue into their corrective action program for resolution. The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and
 
containment) protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier. This finding had a crosscutting aspect in the area of human performance. Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)).
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Current Radiological Information Prior to Entering a High Radiation Area Green. The inspector reviewed a self-revealing, noncited violation of Technical Specification 6.12.1.b that resulted when workers did not obtain current radiological information before entering a high radiation area as required by the Technical Specifications. On December 12, 2006, two workers accessed a high radiation area near the Reactor Coolant Pump 1B Cold Leg through a pathway not discussed with radiation protection and received electronic dose rate alarms. Upon investigation, the licensee determined the workers did not clearly communicate the work scope and the travel path for accessing the work areas; therefore, the workers were not briefed for the radiological conditions of the areas near the Reactor Coolant Pump 1B Cold Leg. The peak dose rates for the two workers were 210 millirem per hour and 361 millirem per hour, respectively. Corrective actions implemented by the licensee were that the workers completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and received additional coaching from radiation protection personnel.
The failure to obtain current radiological information prior to entering a high radiation area is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not obtaining high radiation area dose rates does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self and peer checking, when discussing the work scope and work areas with radiation protection staff [H.4.(a)].
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Green. The inspector reviewed two examples of a self-revealing, noncited violation of Technical Specification 5.4.1 that resulted when workers failed to follow their radiation work permit instructions. The first example occurred on October 11, 2007, when an operator accessed Valves RC 109 and RC 110 by a travel path not discussed with radiation protection personnel and without obtaining current radiological conditions as specified in the radiation work permit.
As the operator passed through the pipe-chase to access the valves, the worker received a dose rate alarm. The highest dose rate levels were 80 millirem per hour along the travel path. The second example occurred on October 12, 2007, when a maintenance mechanic entered the Safeguards B room without a current radiological briefing as specified in the radiation work permit. Radiation protection personnel requested the worker wait to access Safeguards A room while the radiological conditions were changing (shutdown cooling in progress) and did not know the worker also needed to access the B room. The worker, who had previously entered the B room but failed to realize this room
 
also had changing radiological conditions, did not receive current radiological conditions for this room and received a dose rate alarm. The workers peak dose rate was 61 millirem per hour. The licensees corrective actions for the first example were that a radiation protection supervisor conducted an interview with worker, and the worker completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and human performance error review. For the second example, the immediate corrective action was to exclude the individual from the radiological controlled area then perform a human performance error review.
The failure to follow a radiation work permit instruction is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not following their radiation work permit does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self checking, to ensure the full work scope, locations, and radiological conditions were discussed with radiation protection personnel as required by the radiation work permit [H.4.(a)].
Inspection Report# : 2007005 (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 Last modified : August 29, 2008
 
Waterford 3 3Q/2008 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedure Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion V was identified for an inadequate maintenance procedure. Specifically, MM-006-054, Check Valve Inspection (Tilting Disc), lacked sufficient detail to prevent poor workmanship during maintenance on safety injection Tank 1A discharge check Valve SI-335A. This poor workmanship allowed Valve SI-335A to be reassembled with a cocked hinge pin cover, resulting in reactor coolant system (RCS) leakage. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the procedure quality attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because assuming worst case degradation, the Valve SI-335A leak would not result in exceeding the Technical Specification limit for identified RCS leakage. This finding had a crosscutting aspect in the resources component of the human performance area. Specifically, the licensee failed to provide the maintenance technician with a complete and accurate maintenance procedure [H.2(c)].
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Reactor Coolant Pump 1A Seal Leak Green. A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criterion XVI was identified for the licensees failure to promptly identify and correct a significant condition adverse to quality. Specifically, the licensee did not identify a seal leak on reactor coolant Pump 1A in a timely fashion. During efforts to identify the source of leakage, the licensee effectively ruled out the reactor coolant pump seal areas based on an incorrect assumption. When no other significant sources of leakage could be found, the decision was made to monitor the leakage and take no further actions until the mid-cycle outage. This unidentified reactor coolant system leakage caused degradation to the reactor coolant pump cover, main casing stud nuts, shroud wall, and carbon steel flanges. The licensee entered this issue into their corrective action program for resolution.
The finding is more than minor because it challenges the equipment performance attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Appendix A Phase 1 screening worksheet, the issue screened as having very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available. This finding had a crosscutting aspect in the area of human performance associated with decision-making in that the licensee did not use conservative assumptions in the reactor coolant system leakage investigation [H.1(b)].
Inspection Report# : 2007005 (pdf)
Mitigating Systems Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to re-evaluate previously identified boric acid leaks The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions for a condition adverse to quality. Specifically, the licensee developed a corrective action to evaluate the condition of existing boric acid leaks. However, the effort failed to identify and evaluate multiple existing boric acid leaks on safety related components, including some that had deteriorated since initial discovery. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2007 3951.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. Specifically, some
 
unchecked boric acid leaks may have worsened and corroded safety related equipment. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding had very low risk significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because engineers failed to implement proper error prevention techniques when identifying boric acid leaks for additional review H.4 (a).
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to correct "Fuel Oil Receipt and Transfer" procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to implement adequate corrective actions for a deficient emergency diesel generator fuel tank filling procedure (a condition adverse to quality). The licensee had identified the deficiency following a previous event when fuel oil leaked out of multiple fuel oil injectors during a diesel run. Procedural steps were needed to adequately vent the fill line following pressurization during fuel oil tank filling. However, the licensee only corrected the procedure in one section and, when a different section was used, the problem reoccurred. The fuel oil leak led to the emergency diesel generator being declared inoperable. In addition, the fuel oil created a potential fire hazard. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008 1345.
The finding was more than minor because it was similar to nonminor example 4.f in Inspection Manual Chapter 0612, "Examples of Minor Issues," in that emergency diesel generator operability was affected. Further, the oil created a fire hazard. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Essential chiller AB return Header B Isolation Valve CHW 786B Misposition The inspectors documented a self revealing noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to correctly position a valve during a surveillance. The procedure required operators to position the essential Chiller AB return Header B isolation Valve CHW 786B closed but operators left the valve in the open position. This resulted in cross connecting the essential services chilled water Loops A and B, which led to an unplanned entry into Technical Specifications 3.7.12 and 3.0.3. The violation was revealed through a control room alarm. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0778.
The finding was more than minor because, if left uncorrected, would have become a more significant safety concern. Specifically, with both loops of the essential services chilled water system cross connected, the system was no longer single-failure proof. A leak in one of the essential chilled water loops would have caused both units to become inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because operators failed to implement self-checking techniques when performing procedure steps H.4(a).
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation ACCW pump failure due to inaccurate operator aid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to correct a condition adverse to quality (inadequate instructions that led to low fuel oil and the failure of auxiliary component cooling water pump bearing). Specifically, the licensee's corrective action for a previous event called for an operator aid (oil level label). However, the operator aid contained incorrect and confusing information. Consequently, another auxiliary component cooling water pump failed. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0350.
The finding was more than minor because it was similar to nonminor violation example 4.f in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that the problem affected auxiliary component cooling water Pump B operability. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because
 
it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had crosscutting aspects associated with Human Performance area, resources program component, because the licensee failed to have correct labeling on components H.2(c).
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Review Process Green. The inspectors identified a Green noncited violation of Technical Specification 6.8.1.a (Procedures) for failure to correctly implement a procedure recommended in Appendix A of Regulatory Guide 1.33. Specifically, the failure to follow Site Procedure W2.109, Procedure Development, Review, and Approval, led to the unapproved deletion of the Special Scope section of the Quality Assurance Program Manual. The Special Scope section contained the fire protection quality assurance (QA) program components and discussion for their implementation. This deleted information is required by the Waterford 3 Steam Electric Station License Condition 2.C.9. The licensee entered this issue into their corrective action program for resolution.
The finding was more than minor because if left uncorrected, it would become a more significant safety concern. Using Inspection Manual Chapter 0609, Appendix F, this finding can be assigned a low degradation rating and screen as green, since current QA audit standards contain a similar level of detail as the criteria contained in the deleted Special Scope document.
Inspection Report# : 2007005 (pdf)
Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for a Fire in Vital Switchgear Room B The inspectors identified two examples of a noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility. In the first example, the pre-fire strategy for vital switchgear Room B did not contain adequate information regarding the doors required to be open to allow the desired ventilation flowpath, nor did it contain the required number of smoke ejectors necessary to desmoke the switchgear room in a manner that would allow the implementation of OP-901-524, Fire In Areas Affecting Safe Shutdown. In the second example, the licensee did not take corrective actions for a previously identified issue in a timely fashion. Specifically, the deficiencies in the pre-fire strategy for vital switchgear Room B were first identified on August 21, 2006. The deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the non-conformance with licensee management. The licensee entered this deficiency into their corrective action program for resolution. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution. Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)).
Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:        Oct 07, 2007 Identified By: NRC Item Type: NCV NonCited Violation Missed Reactor Coolant System Chemistry Samples The inspectors identified a noncited violation of Technical Specification (TS) 3.4.7 for multiple failures to complete a radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity. Specifically, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the TS required interval of 136 to 229 days. EBAR is the average of the sum of average beta and gamma energies per disintegration for isotopes, other than radioiodines, with half-lives greater than fifteen minutes. Daily RCS samples are compared to this calculated value in order to ensure that 10CFR50.67 dose limits at the site boundary are not exceeded in the event of an accident scenario. The licensee entered this issue into their corrective action program for resolution. The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using the
 
Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier. This finding had a crosscutting aspect in the area of human performance.
Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)).
Inspection Report# : 2007004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        May 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to conspicuously post a radiation area The inspectors identified a NCV of 10 CFR 20.1902 because the licensee failed to post a radiation area conspicuously. On May 14, 2008, the inspectors toured the hot machine shop and noted a box with high radiation area signs attached. Dose rates around the box ranged from 55 to 90 millirems per hour at 30 centimeters. The inspectors noted there was no posting to identify the radiation area. The nearest radiation area posting was on the entry door of the decontamination room, outside the hot machine shop. As a result of the inspectors finding, the licensee erected a rope barricade around the radiation area and posted it conspicuously.
The finding was more than minor because it was associated with one of the cornerstone attributes and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Because the finding involved the potential for unplanned, unintended dose resulting from conditions that were contrary to NRC regulations, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The inspectors determined that the finding had no more than very low safety significance because: (1) it did not involve (ALARA) planning and controls, (2) there was no personnel overexposure, (3) there was no substantial potential for personnel overexposure, and (4) the finding did not compromise the licensees ability to assess dose. The finding also had a cross-cutting aspect in the area of human performance, resource component, because the licensee did not have complete procedures. (H.2.c)
Inspection Report# : 2008003 (pdf)
Significance:        Apr 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure when entering the radiological controlled area The inspectors reviewed two examples of a self-revealing, noncited violation of Technical Specification 6.8.1 because workers failed to follow procedural requirements when preparing to enter the radiological controlled area.
The first example, on April 28, 2008, involved a contract employee who informed the radiation protection shift control technician he would be working on the Reactor Coolant Pump 1B platform where dose rates were below 350 millirems per hour. Subsequently, the contract employee entered another area, one which had not been surveyed and on which the worker had not been briefed, and received a dose rate alarm measuring 553 millirems per hour. The second example, on April 30, 2008, involved a rigger who was assigned to help rig and lift a reactor coolant pump seal from the pump to the top of the D-ring. However, the rigger did not report to radiation protection personnel to receive a briefing on the dose rates in the area of Reactor Coolant Pump 1A. Before being reassigned, the rigger was briefed for an area with dose rates less than 180 millirems per hour, but during his work on the reactor coolant pump, the worker entered an area with dose rates as high as 628 millirems per hour and received a dose rate alarm. Radiation protection personnel counseled the workers and documented the occurrences in the corrective action program.
The occurrence involved the program attributes of exposure control and affected the cornerstone objective, in that the failure of the workers to follow procedural guidance and inform radiation protection personnel of the workers intended activities work area resulted in the workers being unknowledgeable of the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not: (1) an ALARA finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a cross-cutting aspect in the area of human performance, work practices component, because the workers failed to use human error prevention techniques such as self and peer checking. (H.4.a)
Inspection Report# : 2008003 (pdf)
 
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Current Radiological Information Prior to Entering a High Radiation Area Green. The inspector reviewed a self-revealing, noncited violation of Technical Specification 6.12.1.b that resulted when workers did not obtain current radiological information before entering a high radiation area as required by the Technical Specifications. On December 12, 2006, two workers accessed a high radiation area near the Reactor Coolant Pump 1B Cold Leg through a pathway not discussed with radiation protection and received electronic dose rate alarms. Upon investigation, the licensee determined the workers did not clearly communicate the work scope and the travel path for accessing the work areas; therefore, the workers were not briefed for the radiological conditions of the areas near the Reactor Coolant Pump 1B Cold Leg. The peak dose rates for the two workers were 210 millirem per hour and 361 millirem per hour, respectively. Corrective actions implemented by the licensee were that the workers completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and received additional coaching from radiation protection personnel.
The failure to obtain current radiological information prior to entering a high radiation area is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not obtaining high radiation area dose rates does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self and peer checking, when discussing the work scope and work areas with radiation protection staff [H.4.(a)].
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Green. The inspector reviewed two examples of a self-revealing, noncited violation of Technical Specification 5.4.1 that resulted when workers failed to follow their radiation work permit instructions. The first example occurred on October 11, 2007, when an operator accessed Valves RC 109 and RC 110 by a travel path not discussed with radiation protection personnel and without obtaining current radiological conditions as specified in the radiation work permit. As the operator passed through the pipe-chase to access the valves, the worker received a dose rate alarm. The highest dose rate levels were 80 millirem per hour along the travel path. The second example occurred on October 12, 2007, when a maintenance mechanic entered the Safeguards B room without a current radiological briefing as specified in the radiation work permit. Radiation protection personnel requested the worker wait to access Safeguards A room while the radiological conditions were changing (shutdown cooling in progress) and did not know the worker also needed to access the B room. The worker, who had previously entered the B room but failed to realize this room also had changing radiological conditions, did not receive current radiological conditions for this room and received a dose rate alarm. The workers peak dose rate was 61 millirem per hour. The licensees corrective actions for the first example were that a radiation protection supervisor conducted an interview with worker, and the worker completed an electronic alarming dosimeter dose/dose rate alarm questionnaire and human performance error review. For the second example, the immediate corrective action was to exclude the individual from the radiological controlled area then perform a human performance error review.
The failure to follow a radiation work permit instruction is a performance deficiency. This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affected the Occupational Radiation Safety cornerstone objective, in that workers not following their radiation work permit does not ensure adequate protection of the worker health and safety from additional personal exposure. The finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Further, this finding had a human performance cross-cutting aspect in the work practices component because the workers did not use human error prevention techniques, such as self checking, to ensure the full work scope, locations, and radiological conditions were discussed with radiation protection personnel as required by the radiation work permit [H.4.(a)].
Inspection Report# : 2007005 (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 Last modified : November 26, 2008
 
Waterford 3 4Q/2008 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Test Procedure for Safety Injection Valves S1-405A(B) Post Modification Testing Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion III due to the failure by the licensee to perform adequate post modification testing to evaluate the adequacy of design modifications made to the actuators of low pressure safety injection Isolation Valves SI-405A(B). This led to the licensee failing to identify a fundamental difference in the manner that the air operated valve actuator operated resulting in the valve popping open instead of slowly opening, creating a pressure transient that resulted in the lifting of the low temperature overpressure relief valve causing an intersystem loss-of-coolant event. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-4161.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both Valves SI-405A(B) inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss-of-inventory event. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. This finding had a crosscutting aspect in Human Performance, specifically the Resources aspect [H.2(a)] because the licensee failed to maintain adequate design margins. Specifically, the licensees pneumatic actuator for SI-405B could not overcome the pressure locking mechanism until twelve minutes into a fifteen minute time limit, after receiving the open demand signal. This led to the instantaneous valve disc displacement when the valve popped open causing the pressure surge, which resulted in the opening of relief valve SI-406B and subsequent loss of inventory event (Section 4OA5).
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions
* Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI due to the failure by the licensee to take prompt corrective actions following the identification of an inadequate testing method used for determining the integrity of the Essential Chiller B heat exchanger tubing. Failure to take this timely action resulted in an inadvertent tube rupture and inoperability of Essential Chiller B. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5342.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because, assuming worst case degradation of both the B and AB Essential Chillers failing, the redundant A Essential Chiller would still have been available for accident mitigation. This finding had a crosscutting aspect in Problem Identification and Resolution, specifically the Corrective Action Program aspect [P.1
 
(d)] because the licensee failed to take appropriate corrective actions to address a degrading condition in a timely manner. Specifically, the failure to perform timely tube inspections of Essential Chiller B, following the identification of an inadequate testing methodology used for identifying Essential Chiller heat exchanger tubing degradation (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Essential Chiller AB Component Failure Due to Inadequate Procedural Guidance
* Green. The inspectors reviewed a self revealing noncited violation of Technical Specification 6.8.1.a for failure to provide documented instructions appropriate to the circumstances as recommended in Appendix A of Regulatory Guide 1.33. The failure by the licensee to provide adequate guidance for the replacement of the Essential Chiller AB compressor motor temperature sensor resulted in the reintroduction of a failure mechanism that had previously been corrected. This subsequently led to the failure of the temperature sensor wiring and inoperability of Essential Chiller AB. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5471.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because the redundant Essential Chillers A and B would still have been available for accident mitigation. Based on the guidance provided in Manual Chapter 0612, Appendix B, Section 1-5, Screen for Cross-Cutting Aspects, this finding did not have a crosscutting aspect because it was not considered to be reflective of current licensee performance. Specifically, the licensees failure to update the model work instructions in 2000 was a latent issue, whereby the licensee did not have a reasonable opportunity to identify the problem prior to August, 2008. In addition, the licensee has since instituted programs and processes such that the problem would not reasonably occur today (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calculations Used for Operability determination of SI-405 A(B)
Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III to address three examples of inadequate calculations involving shutdown cooling Valves SI-405A and SI-405B. The calculations were also used, in part, to support valve operability, which made an existing operability assessment invalid. First, a calculation performed by a contractor to estimate the bounding thrust requirements for pressure locking contained errors and used mathematical formulas out of their intended context without applying uncertainties to account for the differences. Recent operational experience with these valves was inconsistent with the calculation's conclusions. In addition, the licensee failed to meet their quality assurance program requirements that specified that engineers perform a design verification of the calculation prior to use. Second, the licensee's calculation, that demonstrated valve actuator thrust capabilities, contained errors. Specifically, it failed to account for the friction between the actuator piston disk and walls as well as the weight of components. Third, a calculation that determined that the temperature within the valve bonnet would not heat up during small break loss of coolant accidents and faulted steam generator accidents was inadequate, in that it failed to address a faulted steam generator event, it used heat transfer calculation methods on water that were intended only for solid materials, it failed to model all components, and it failed to determine the temperatures inside the valve bonnets, which was the overriding variable of interest. The licensee entered the finding into the corrective action program as Condition Report CR-WF3-2009-00127.
This finding was more than minor because it was similar to non-minor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt concerning the operability of Valves SI-405A(B). The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown
 
Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI-405A(B) valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high-pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution (P.1(c)) because engineers failed to thoroughly evaluate the potential for valve pressure-locking. The calculations were completed in 2008 and were indicative of current performance (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate pressure locking calculation Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III (Design Control) for an inadequate "pressure locking" design calculation for shutdown cooling Valves SI-405A and SI-405B. Plant engineers also used the calculation to support valve operability following a valve malfunction, which appeared to be caused by pressure locking. Entergy engineers had derived valve bonnet leakage rates (for pressure locking conditions) from local leak rate testing results. However, a national laboratory had already proven the Entergy theory invalid and plant engineers had taken no steps to validate the theory themselves. Finally, in response to an NRC generic letter concerning pressure locking and thermal binding of valves, the licensee engineers' conclusions were based on incorrect facts and improper assumptions. Licensee personnel entered the noncited violation into the corrective action program as Condition Report CR WF3 2008 4292.
The failures to perform: (1) an adequate engineering calculation and (2) a valid operability determination were performance deficiencies. This finding was more than minor because it was similar to nonminor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that, there was a reasonable doubt concerning the operability of Valves SI-405A/B. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI 405A/B valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution [P.1 (c)] because engineers failed to thoroughly evaluate the potential for valve pressure locking. The calculation was completed in 2008 and was indicative of current performance.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow integrated EDG test procedure Green. The inspectors identified a noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to open the Train A low pressure safety injection pump suction valve prior to pump operation during a surveillance. The butterfly valve was installed 90 degrees out of position and was closed when operators believed it was open. After starting the pump, operators observed loud noises coming from the unit and secured it 8 minutes later. Pump operation without adequate net positive suction head could cause damage. The valve's postmaintenance test was scheduled after the noted surveillance test, and the surveillance was not intended to check the valve's function. The safety injection train was considered inoperable but available at the time. Licensee personnel entered the noncited violation into the corrective action program as Condition Reports CR-WF3-2008-2280 and CR-WF3-2008-3045.
 
This finding was more than minor because it affected both the configuration control and the equipment performance attributes of the Mitigating Systems Cornerstone objective to ensure reliability of the low pressure safety injection system. In addition, this condition, if left uncorrected, would also become a more significant safety concern.
Equipment could be damaged without adequate postmaintenance checks prior to operation. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding was of very low risk significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had a crosscutting aspect in the area of human performance, associated with the decision-making component, in that, the plant personnel used nonconservative assumptions and chose to use the pump suction valve for system operation prior to verifying that the valve was properly assembled [H.1(b)]
Inspection Report# : 2008004 (pdf)
Significance:      Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to re-evaluate previously identified boric acid leaks The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions for a condition adverse to quality. Specifically, the licensee developed a corrective action to evaluate the condition of existing boric acid leaks. However, the effort failed to identify and evaluate multiple existing boric acid leaks on safety related components, including some that had deteriorated since initial discovery. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2007 3951.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. Specifically, some unchecked boric acid leaks may have worsened and corroded safety related equipment.
Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding had very low risk significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because engineers failed to implement proper error prevention techniques when identifying boric acid leaks for additional review H.4 (a).
Inspection Report# : 2008002 (pdf)
Significance:      Apr 07, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to correct "Fuel Oil Receipt and Transfer" procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to implement adequate corrective actions for a deficient emergency diesel generator fuel tank filling procedure (a condition adverse to quality). The licensee had identified the deficiency following a previous event when fuel oil leaked out of multiple fuel oil injectors during a diesel run. Procedural steps were needed to adequately vent the fill line following pressurization during fuel oil tank filling. However, the licensee only corrected the procedure in one section and, when a different section was used, the problem reoccurred. The fuel oil leak led to the emergency diesel generator being declared inoperable. In addition, the fuel oil created a potential fire hazard. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008 1345.
The finding was more than minor because it was similar to nonminor example 4.f in Inspection Manual Chapter 0612, "Examples of Minor Issues," in that emergency diesel generator operability was affected. Further, the oil created a fire hazard. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2)
 
represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Essential chiller AB return Header B Isolation Valve CHW 786B Misposition The inspectors documented a self revealing noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to correctly position a valve during a surveillance. The procedure required operators to position the essential Chiller AB return Header B isolation Valve CHW 786B closed but operators left the valve in the open position. This resulted in cross connecting the essential services chilled water Loops A and B, which led to an unplanned entry into Technical Specifications 3.7.12 and 3.0.3. The violation was revealed through a control room alarm. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0778.
The finding was more than minor because, if left uncorrected, would have become a more significant safety concern.
Specifically, with both loops of the essential services chilled water system cross connected, the system was no longer single-failure proof. A leak in one of the essential chilled water loops would have caused both units to become inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because operators failed to implement self-checking techniques when performing procedure steps H.4(a).
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation ACCW pump failure due to inaccurate operator aid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to correct a condition adverse to quality (inadequate instructions that led to low fuel oil and the failure of auxiliary component cooling water pump bearing). Specifically, the licensee's corrective action for a previous event called for an operator aid (oil level label). However, the operator aid contained incorrect and confusing information. Consequently, another auxiliary component cooling water pump failed. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0350.
The finding was more than minor because it was similar to nonminor violation example 4.f in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that the problem affected auxiliary component cooling water Pump B operability. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had crosscutting aspects associated with Human Performance area, resources program component, because the licensee failed to have correct labeling on components H.2(c).
Inspection Report# : 2008002 (pdf)
Barrier Integrity
 
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct a condition adverse to quality.
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to promptly identify and correct a condition adverse to quality. Specifically, from March 20, 2007, through October 27, 2008, personnel failed to identify and correct a condition, which allowed containment vacuum relief valve differential pressure switches to operate in pressures that exceeded the designed operating pressure of the switches. The licensee implemented interim corrective actions to ensure operability. Specifically, the licensee increased the test frequency and adjusted the switches to reduce the effects of the deficient condition. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05106.
The performance deficiency associated with this finding involved the failure to promptly identify and correct a condition adverse to quality that could affect containment integrity. This finding was greater than minor because it affected the Configuration Control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that the containment physical design barrier protected the public from radionuclide releases caused by an event. Using the NRC Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. This finding had a crosscutting aspect in the area of human performance, associated with the decision making component, in that, licensee personnel failed to make conservative decisions related to equipment operation in accordance with design requirements (H.1(b)).
Inspection Report# : 2008007 (pdf)
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a pressure boundary valve The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, for a failure to follow procedure, when the licensee failed to complete an adequate operability evaluation for Valve SI 142A. Specifically, on August 21, 2008, the licensee failed to follow Procedure EN OP 104, "Operability Determinations," Revision 3, because personnel did not determine the leak rate solely through the required pressure boundary valve. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05077.
The failure to perform an adequate operability evaluation on safety related plant equipment in accordance with Procedure EN OP 104 is a performance deficiency. The team determined this finding was greater than minor from review of Manual Chapter 0612, Appendix E, "Examples of Minor Issues." The finding was similar to non minor finding Example 3.j in that reasonable doubt existed related to the operability of Valve SI 142A. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. The finding had a crosscutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate valve operability (P.1(c)).
Inspection Report# : 2008007 (pdf)
Emergency Preparedness Occupational Radiation Safety
 
Significance:      May 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to conspicuously post a radiation area The inspectors identified a NCV of 10 CFR 20.1902 because the licensee failed to post a radiation area conspicuously.
On May 14, 2008, the inspectors toured the hot machine shop and noted a box with high radiation area signs attached.
Dose rates around the box ranged from 55 to 90 millirems per hour at 30 centimeters. The inspectors noted there was no posting to identify the radiation area. The nearest radiation area posting was on the entry door of the decontamination room, outside the hot machine shop. As a result of the inspectors finding, the licensee erected a rope barricade around the radiation area and posted it conspicuously.
The finding was more than minor because it was associated with one of the cornerstone attributes and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Because the finding involved the potential for unplanned, unintended dose resulting from conditions that were contrary to NRC regulations, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The inspectors determined that the finding had no more than very low safety significance because: (1) it did not involve (ALARA) planning and controls, (2) there was no personnel overexposure, (3) there was no substantial potential for personnel overexposure, and (4) the finding did not compromise the licensees ability to assess dose. The finding also had a cross-cutting aspect in the area of human performance, resource component, because the licensee did not have complete procedures. (H.2.c)
Inspection Report# : 2008003 (pdf)
Significance:      Apr 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure when entering the radiological controlled area The inspectors reviewed two examples of a self-revealing, noncited violation of Technical Specification 6.8.1 because workers failed to follow procedural requirements when preparing to enter the radiological controlled area.
The first example, on April 28, 2008, involved a contract employee who informed the radiation protection shift control technician he would be working on the Reactor Coolant Pump 1B platform where dose rates were below 350 millirems per hour. Subsequently, the contract employee entered another area, one which had not been surveyed and on which the worker had not been briefed, and received a dose rate alarm measuring 553 millirems per hour. The second example, on April 30, 2008, involved a rigger who was assigned to help rig and lift a reactor coolant pump seal from the pump to the top of the D-ring. However, the rigger did not report to radiation protection personnel to receive a briefing on the dose rates in the area of Reactor Coolant Pump 1A. Before being reassigned, the rigger was briefed for an area with dose rates less than 180 millirems per hour, but during his work on the reactor coolant pump, the worker entered an area with dose rates as high as 628 millirems per hour and received a dose rate alarm. Radiation protection personnel counseled the workers and documented the occurrences in the corrective action program.
The occurrence involved the program attributes of exposure control and affected the cornerstone objective, in that the failure of the workers to follow procedural guidance and inform radiation protection personnel of the workers intended activities work area resulted in the workers being unknowledgeable of the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not: (1) an ALARA finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a cross-cutting aspect in the area of human performance, work practices component, because the workers failed to use human error prevention techniques such as self and peer checking. (H.4.a)
Inspection Report# : 2008003 (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 Last modified : April 07, 2009
 
Waterford 3 1Q/2009 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Test Procedure for Safety Injection Valves S1-405A(B) Post Modification Testing Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion III due to the failure by the licensee to perform adequate post modification testing to evaluate the adequacy of design modifications made to the actuators of low pressure safety injection Isolation Valves SI-405A(B). This led to the licensee failing to identify a fundamental difference in the manner that the air operated valve actuator operated resulting in the valve popping open instead of slowly opening, creating a pressure transient that resulted in the lifting of the low temperature overpressure relief valve causing an intersystem loss-of-coolant event. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-4161.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both Valves SI-405A(B) inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss-of-inventory event. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. This finding had a crosscutting aspect in Human Performance, specifically the Resources aspect [H.2(a)] because the licensee failed to maintain adequate design margins. Specifically, the licensees pneumatic actuator for SI-405B could not overcome the pressure locking mechanism until twelve minutes into a fifteen minute time limit, after receiving the open demand signal. This led to the instantaneous valve disc displacement when the valve popped open causing the pressure surge, which resulted in the opening of relief valve SI-406B and subsequent loss of inventory event (Section 4OA5).
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions
* Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI due to the failure by the licensee to take prompt corrective actions following the identification of an inadequate testing method used for determining the integrity of the Essential Chiller B heat exchanger tubing. Failure to take this timely action resulted in an inadvertent tube rupture and inoperability of Essential Chiller B. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5342.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because, assuming worst case degradation of both the B and AB Essential Chillers failing, the redundant A Essential Chiller would still have been available for accident mitigation. This finding had a crosscutting aspect in Problem Identification and Resolution, specifically the Corrective Action Program aspect [P.1
 
(d)] because the licensee failed to take appropriate corrective actions to address a degrading condition in a timely manner. Specifically, the failure to perform timely tube inspections of Essential Chiller B, following the identification of an inadequate testing methodology used for identifying Essential Chiller heat exchanger tubing degradation (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Essential Chiller AB Component Failure Due to Inadequate Procedural Guidance
* Green. The inspectors reviewed a self revealing noncited violation of Technical Specification 6.8.1.a for failure to provide documented instructions appropriate to the circumstances as recommended in Appendix A of Regulatory Guide 1.33. The failure by the licensee to provide adequate guidance for the replacement of the Essential Chiller AB compressor motor temperature sensor resulted in the reintroduction of a failure mechanism that had previously been corrected. This subsequently led to the failure of the temperature sensor wiring and inoperability of Essential Chiller AB. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5471.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because the redundant Essential Chillers A and B would still have been available for accident mitigation. Based on the guidance provided in Manual Chapter 0612, Appendix B, Section 1-5, Screen for Cross-Cutting Aspects, this finding did not have a crosscutting aspect because it was not considered to be reflective of current licensee performance. Specifically, the licensees failure to update the model work instructions in 2000 was a latent issue, whereby the licensee did not have a reasonable opportunity to identify the problem prior to August, 2008. In addition, the licensee has since instituted programs and processes such that the problem would not reasonably occur today (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calculations Used for Operability determination of SI-405 A(B)
Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III to address three examples of inadequate calculations involving shutdown cooling Valves SI-405A and SI-405B. The calculations were also used, in part, to support valve operability, which made an existing operability assessment invalid. First, a calculation performed by a contractor to estimate the bounding thrust requirements for pressure locking contained errors and used mathematical formulas out of their intended context without applying uncertainties to account for the differences. Recent operational experience with these valves was inconsistent with the calculation's conclusions. In addition, the licensee failed to meet their quality assurance program requirements that specified that engineers perform a design verification of the calculation prior to use. Second, the licensee's calculation, that demonstrated valve actuator thrust capabilities, contained errors. Specifically, it failed to account for the friction between the actuator piston disk and walls as well as the weight of components. Third, a calculation that determined that the temperature within the valve bonnet would not heat up during small break loss of coolant accidents and faulted steam generator accidents was inadequate, in that it failed to address a faulted steam generator event, it used heat transfer calculation methods on water that were intended only for solid materials, it failed to model all components, and it failed to determine the temperatures inside the valve bonnets, which was the overriding variable of interest. The licensee entered the finding into the corrective action program as Condition Report CR-WF3-2009-00127.
This finding was more than minor because it was similar to non-minor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt concerning the operability of Valves SI-405A(B). The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown
 
Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI-405A(B) valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high-pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution (P.1(c)) because engineers failed to thoroughly evaluate the potential for valve pressure-locking. The calculations were completed in 2008 and were indicative of current performance (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate pressure locking calculation Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III (Design Control) for an inadequate "pressure locking" design calculation for shutdown cooling Valves SI-405A and SI-405B. Plant engineers also used the calculation to support valve operability following a valve malfunction, which appeared to be caused by pressure locking. Entergy engineers had derived valve bonnet leakage rates (for pressure locking conditions) from local leak rate testing results. However, a national laboratory had already proven the Entergy theory invalid and plant engineers had taken no steps to validate the theory themselves. Finally, in response to an NRC generic letter concerning pressure locking and thermal binding of valves, the licensee engineers' conclusions were based on incorrect facts and improper assumptions. Licensee personnel entered the noncited violation into the corrective action program as Condition Report CR WF3 2008 4292.
The failures to perform: (1) an adequate engineering calculation and (2) a valid operability determination were performance deficiencies. This finding was more than minor because it was similar to nonminor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that, there was a reasonable doubt concerning the operability of Valves SI-405A/B. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI 405A/B valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution [P.1 (c)] because engineers failed to thoroughly evaluate the potential for valve pressure locking. The calculation was completed in 2008 and was indicative of current performance.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow integrated EDG test procedure Green. The inspectors identified a noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to open the Train A low pressure safety injection pump suction valve prior to pump operation during a surveillance. The butterfly valve was installed 90 degrees out of position and was closed when operators believed it was open. After starting the pump, operators observed loud noises coming from the unit and secured it 8 minutes later. Pump operation without adequate net positive suction head could cause damage. The valve's postmaintenance test was scheduled after the noted surveillance test, and the surveillance was not intended to check the valve's function. The safety injection train was considered inoperable but available at the time. Licensee personnel entered the noncited violation into the corrective action program as Condition Reports CR-WF3-2008-2280 and CR-WF3-2008-3045.
 
This finding was more than minor because it affected both the configuration control and the equipment performance attributes of the Mitigating Systems Cornerstone objective to ensure reliability of the low pressure safety injection system. In addition, this condition, if left uncorrected, would also become a more significant safety concern.
Equipment could be damaged without adequate postmaintenance checks prior to operation. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding was of very low risk significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had a crosscutting aspect in the area of human performance, associated with the decision-making component, in that, the plant personnel used nonconservative assumptions and chose to use the pump suction valve for system operation prior to verifying that the valve was properly assembled [H.1(b)]
Inspection Report# : 2008004 (pdf)
Significance:      Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to re-evaluate previously identified boric acid leaks The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions for a condition adverse to quality. Specifically, the licensee developed a corrective action to evaluate the condition of existing boric acid leaks. However, the effort failed to identify and evaluate multiple existing boric acid leaks on safety related components, including some that had deteriorated since initial discovery. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2007 3951.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. Specifically, some unchecked boric acid leaks may have worsened and corroded safety related equipment.
Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding had very low risk significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because engineers failed to implement proper error prevention techniques when identifying boric acid leaks for additional review H.4 (a).
Inspection Report# : 2008002 (pdf)
Significance:      Apr 07, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to correct "Fuel Oil Receipt and Transfer" procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to implement adequate corrective actions for a deficient emergency diesel generator fuel tank filling procedure (a condition adverse to quality). The licensee had identified the deficiency following a previous event when fuel oil leaked out of multiple fuel oil injectors during a diesel run. Procedural steps were needed to adequately vent the fill line following pressurization during fuel oil tank filling. However, the licensee only corrected the procedure in one section and, when a different section was used, the problem reoccurred. The fuel oil leak led to the emergency diesel generator being declared inoperable. In addition, the fuel oil created a potential fire hazard. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008 1345.
The finding was more than minor because it was similar to nonminor example 4.f in Inspection Manual Chapter 0612, "Examples of Minor Issues," in that emergency diesel generator operability was affected. Further, the oil created a fire hazard. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2)
 
represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation Essential chiller AB return Header B Isolation Valve CHW 786B Misposition The inspectors documented a self revealing noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to correctly position a valve during a surveillance. The procedure required operators to position the essential Chiller AB return Header B isolation Valve CHW 786B closed but operators left the valve in the open position. This resulted in cross connecting the essential services chilled water Loops A and B, which led to an unplanned entry into Technical Specifications 3.7.12 and 3.0.3. The violation was revealed through a control room alarm. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0778.
The finding was more than minor because, if left uncorrected, would have become a more significant safety concern.
Specifically, with both loops of the essential services chilled water system cross connected, the system was no longer single-failure proof. A leak in one of the essential chilled water loops would have caused both units to become inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessments." This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because operators failed to implement self-checking techniques when performing procedure steps H.4(a).
Inspection Report# : 2008002 (pdf)
Significance:        Apr 07, 2008 Identified By: NRC Item Type: NCV NonCited Violation ACCW pump failure due to inaccurate operator aid The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to correct a condition adverse to quality (inadequate instructions that led to low fuel oil and the failure of auxiliary component cooling water pump bearing). Specifically, the licensee's corrective action for a previous event called for an operator aid (oil level label). However, the operator aid contained incorrect and confusing information. Consequently, another auxiliary component cooling water pump failed. The licensee entered this deficiency into their corrective action program as Condition Report CR WF3 2008-0350.
The finding was more than minor because it was similar to nonminor violation example 4.f in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that the problem affected auxiliary component cooling water Pump B operability. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had crosscutting aspects associated with Human Performance area, resources program component, because the licensee failed to have correct labeling on components H.2(c).
Inspection Report# : 2008002 (pdf)
Barrier Integrity
 
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct a condition adverse to quality.
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to promptly identify and correct a condition adverse to quality. Specifically, from March 20, 2007, through October 27, 2008, personnel failed to identify and correct a condition, which allowed containment vacuum relief valve differential pressure switches to operate in pressures that exceeded the designed operating pressure of the switches. The licensee implemented interim corrective actions to ensure operability. Specifically, the licensee increased the test frequency and adjusted the switches to reduce the effects of the deficient condition. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05106.
The performance deficiency associated with this finding involved the failure to promptly identify and correct a condition adverse to quality that could affect containment integrity. This finding was greater than minor because it affected the Configuration Control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that the containment physical design barrier protected the public from radionuclide releases caused by an event. Using the NRC Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. This finding had a crosscutting aspect in the area of human performance, associated with the decision making component, in that, licensee personnel failed to make conservative decisions related to equipment operation in accordance with design requirements (H.1(b)).
Inspection Report# : 2008007 (pdf)
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a pressure boundary valve The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, for a failure to follow procedure, when the licensee failed to complete an adequate operability evaluation for Valve SI 142A. Specifically, on August 21, 2008, the licensee failed to follow Procedure EN OP 104, "Operability Determinations," Revision 3, because personnel did not determine the leak rate solely through the required pressure boundary valve. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05077.
The failure to perform an adequate operability evaluation on safety related plant equipment in accordance with Procedure EN OP 104 is a performance deficiency. The team determined this finding was greater than minor from review of Manual Chapter 0612, Appendix E, "Examples of Minor Issues." The finding was similar to non minor finding Example 3.j in that reasonable doubt existed related to the operability of Valve SI 142A. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. The finding had a crosscutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate valve operability (P.1(c)).
Inspection Report# : 2008007 (pdf)
Emergency Preparedness Occupational Radiation Safety
 
Significance:      May 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to conspicuously post a radiation area The inspectors identified a NCV of 10 CFR 20.1902 because the licensee failed to post a radiation area conspicuously.
On May 14, 2008, the inspectors toured the hot machine shop and noted a box with high radiation area signs attached.
Dose rates around the box ranged from 55 to 90 millirems per hour at 30 centimeters. The inspectors noted there was no posting to identify the radiation area. The nearest radiation area posting was on the entry door of the decontamination room, outside the hot machine shop. As a result of the inspectors finding, the licensee erected a rope barricade around the radiation area and posted it conspicuously.
The finding was more than minor because it was associated with one of the cornerstone attributes and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Because the finding involved the potential for unplanned, unintended dose resulting from conditions that were contrary to NRC regulations, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The inspectors determined that the finding had no more than very low safety significance because: (1) it did not involve (ALARA) planning and controls, (2) there was no personnel overexposure, (3) there was no substantial potential for personnel overexposure, and (4) the finding did not compromise the licensees ability to assess dose. The finding also had a cross-cutting aspect in the area of human performance, resource component, because the licensee did not have complete procedures. (H.2.c)
Inspection Report# : 2008003 (pdf)
Significance:      Apr 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure when entering the radiological controlled area The inspectors reviewed two examples of a self-revealing, noncited violation of Technical Specification 6.8.1 because workers failed to follow procedural requirements when preparing to enter the radiological controlled area.
The first example, on April 28, 2008, involved a contract employee who informed the radiation protection shift control technician he would be working on the Reactor Coolant Pump 1B platform where dose rates were below 350 millirems per hour. Subsequently, the contract employee entered another area, one which had not been surveyed and on which the worker had not been briefed, and received a dose rate alarm measuring 553 millirems per hour. The second example, on April 30, 2008, involved a rigger who was assigned to help rig and lift a reactor coolant pump seal from the pump to the top of the D-ring. However, the rigger did not report to radiation protection personnel to receive a briefing on the dose rates in the area of Reactor Coolant Pump 1A. Before being reassigned, the rigger was briefed for an area with dose rates less than 180 millirems per hour, but during his work on the reactor coolant pump, the worker entered an area with dose rates as high as 628 millirems per hour and received a dose rate alarm. Radiation protection personnel counseled the workers and documented the occurrences in the corrective action program.
The occurrence involved the program attributes of exposure control and affected the cornerstone objective, in that the failure of the workers to follow procedural guidance and inform radiation protection personnel of the workers intended activities work area resulted in the workers being unknowledgeable of the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not: (1) an ALARA finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a cross-cutting aspect in the area of human performance, work practices component, because the workers failed to use human error prevention techniques such as self and peer checking. (H.4.a)
Inspection Report# : 2008003 (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 Last modified : June 05, 2009
 
Waterford 3 2Q/2009 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Test Procedure for Safety Injection Valves S1-405A(B) Post Modification Testing Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion III due to the failure by the licensee to perform adequate post modification testing to evaluate the adequacy of design modifications made to the actuators of low pressure safety injection Isolation Valves SI-405A(B). This led to the licensee failing to identify a fundamental difference in the manner that the air operated valve actuator operated resulting in the valve popping open instead of slowly opening, creating a pressure transient that resulted in the lifting of the low temperature overpressure relief valve causing an intersystem loss-of-coolant event. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-4161.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both Valves SI-405A(B) inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss-of-inventory event. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. This finding had a crosscutting aspect in Human Performance, specifically the Resources aspect [H.2(a)] because the licensee failed to maintain adequate design margins. Specifically, the licensees pneumatic actuator for SI-405B could not overcome the pressure locking mechanism until twelve minutes into a fifteen minute time limit, after receiving the open demand signal. This led to the instantaneous valve disc displacement when the valve popped open causing the pressure surge, which resulted in the opening of relief valve SI-406B and subsequent loss of inventory event (Section 4OA5).
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:      Apr 07, 2009 Identified By: NRC Item Type: FIN Finding Failure to Follow Commitment Tracking Procedures The inspectors identified a finding because the licensee inadvertently deleted procedural steps to recover an emergency diesel generator during a severe accident. The steps were part of a formal commitment to the NRC. The licensee had failed to follow the site commitment management program when making the procedure change and the procedure writer failed to understand the basis for the steps prior to deleting them. The licensee entered this finding in their corrective action program as Condition Reports CR WF3-2009-0193 and CR WF3-2009-1616.
The finding was more than minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, during a severe accident, operators would not have an appropriate mitigation strategy for starting an emergency diesel generator under certain severe accident conditions. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because the finding: (1) could result in a loss of functionality of an emergency diesel generator; (2) did not represent a
 
loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of Human Performance, Decision Making component [H.1(a)], because the licensee failed to use a systematic process when removing the procedural steps Inspection Report# : 2009002 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Voltage Readings Following a Single Cell Battery Charge The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V (Instructions, Procedures and Drawings) because the licensee failed to implement instructions that were intended to help troubleshoot a defective 125 Vdc battery cell. In response to the degraded cell, the licensee had established additional measures to monitor the cell following charging to ensure proper cell operation. However, the licensee did not perform the monitoring. Once identified by the inspectors, the licensee performed more frequent cell tests. The licensee subsequently replaced the faulty cell. The licensee entered this finding into their corrective action program as Condition Reports CR-WF3-2009-1088 and CR-WF3-2009-1099.
The finding was more than minor because it could have resulted in a more significant safety concern if left uncorrected. Specifically, the normal monitoring period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality [P.1(d)]
Inspection Report# : 2009002 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions
* Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI due to the failure by the licensee to take prompt corrective actions following the identification of an inadequate testing method used for determining the integrity of the Essential Chiller B heat exchanger tubing. Failure to take this timely action resulted in an inadvertent tube rupture and inoperability of Essential Chiller B. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5342.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because, assuming worst case degradation of both the B and AB Essential Chillers failing, the redundant A Essential Chiller would still have been available for accident mitigation. This finding had a crosscutting aspect in Problem Identification and Resolution, specifically the Corrective Action Program aspect [P.1 (d)] because the licensee failed to take appropriate corrective actions to address a degrading condition in a timely manner. Specifically, the failure to perform timely tube inspections of Essential Chiller B, following the identification of an inadequate testing methodology used for identifying Essential Chiller heat exchanger tubing degradation (Section 4OA2).
Inspection Report# : 2008005 (pdf)
 
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Essential Chiller AB Component Failure Due to Inadequate Procedural Guidance
* Green. The inspectors reviewed a self revealing noncited violation of Technical Specification 6.8.1.a for failure to provide documented instructions appropriate to the circumstances as recommended in Appendix A of Regulatory Guide 1.33. The failure by the licensee to provide adequate guidance for the replacement of the Essential Chiller AB compressor motor temperature sensor resulted in the reintroduction of a failure mechanism that had previously been corrected. This subsequently led to the failure of the temperature sensor wiring and inoperability of Essential Chiller AB. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5471.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because the redundant Essential Chillers A and B would still have been available for accident mitigation. Based on the guidance provided in Manual Chapter 0612, Appendix B, Section 1-5, Screen for Cross-Cutting Aspects, this finding did not have a crosscutting aspect because it was not considered to be reflective of current licensee performance. Specifically, the licensees failure to update the model work instructions in 2000 was a latent issue, whereby the licensee did not have a reasonable opportunity to identify the problem prior to August, 2008. In addition, the licensee has since instituted programs and processes such that the problem would not reasonably occur today (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calculations Used for Operability determination of SI-405 A(B)
Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III to address three examples of inadequate calculations involving shutdown cooling Valves SI-405A and SI-405B. The calculations were also used, in part, to support valve operability, which made an existing operability assessment invalid. First, a calculation performed by a contractor to estimate the bounding thrust requirements for pressure locking contained errors and used mathematical formulas out of their intended context without applying uncertainties to account for the differences. Recent operational experience with these valves was inconsistent with the calculation's conclusions. In addition, the licensee failed to meet their quality assurance program requirements that specified that engineers perform a design verification of the calculation prior to use. Second, the licensee's calculation, that demonstrated valve actuator thrust capabilities, contained errors. Specifically, it failed to account for the friction between the actuator piston disk and walls as well as the weight of components. Third, a calculation that determined that the temperature within the valve bonnet would not heat up during small break loss of coolant accidents and faulted steam generator accidents was inadequate, in that it failed to address a faulted steam generator event, it used heat transfer calculation methods on water that were intended only for solid materials, it failed to model all components, and it failed to determine the temperatures inside the valve bonnets, which was the overriding variable of interest. The licensee entered the finding into the corrective action program as Condition Report CR-WF3-2009-00127.
This finding was more than minor because it was similar to non-minor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt concerning the operability of Valves SI-405A(B). The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI-405A(B) valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high-pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution (P.1(c)) because
 
engineers failed to thoroughly evaluate the potential for valve pressure-locking. The calculations were completed in 2008 and were indicative of current performance (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate pressure locking calculation Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III (Design Control) for an inadequate "pressure locking" design calculation for shutdown cooling Valves SI-405A and SI-405B. Plant engineers also used the calculation to support valve operability following a valve malfunction, which appeared to be caused by pressure locking. Entergy engineers had derived valve bonnet leakage rates (for pressure locking conditions) from local leak rate testing results. However, a national laboratory had already proven the Entergy theory invalid and plant engineers had taken no steps to validate the theory themselves. Finally, in response to an NRC generic letter concerning pressure locking and thermal binding of valves, the licensee engineers' conclusions were based on incorrect facts and improper assumptions. Licensee personnel entered the noncited violation into the corrective action program as Condition Report CR WF3 2008 4292.
The failures to perform: (1) an adequate engineering calculation and (2) a valid operability determination were performance deficiencies. This finding was more than minor because it was similar to nonminor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that, there was a reasonable doubt concerning the operability of Valves SI-405A/B. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI 405A/B valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution [P.1 (c)] because engineers failed to thoroughly evaluate the potential for valve pressure locking. The calculation was completed in 2008 and was indicative of current performance.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 16, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow integrated EDG test procedure Green. The inspectors identified a noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to open the Train A low pressure safety injection pump suction valve prior to pump operation during a surveillance. The butterfly valve was installed 90 degrees out of position and was closed when operators believed it was open. After starting the pump, operators observed loud noises coming from the unit and secured it 8 minutes later. Pump operation without adequate net positive suction head could cause damage. The valve's postmaintenance test was scheduled after the noted surveillance test, and the surveillance was not intended to check the valve's function. The safety injection train was considered inoperable but available at the time. Licensee personnel entered the noncited violation into the corrective action program as Condition Reports CR-WF3-2008-2280 and CR-WF3-2008-3045.
This finding was more than minor because it affected both the configuration control and the equipment performance attributes of the Mitigating Systems Cornerstone objective to ensure reliability of the low pressure safety injection system. In addition, this condition, if left uncorrected, would also become a more significant safety concern.
Equipment could be damaged without adequate postmaintenance checks prior to operation. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the finding was of very low risk significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of
 
equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding had a crosscutting aspect in the area of human performance, associated with the decision-making component, in that, the plant personnel used nonconservative assumptions and chose to use the pump suction valve for system operation prior to verifying that the valve was properly assembled [H.1(b)]
Inspection Report# : 2008004 (pdf)
Barrier Integrity Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct a condition adverse to quality.
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to promptly identify and correct a condition adverse to quality. Specifically, from March 20, 2007, through October 27, 2008, personnel failed to identify and correct a condition, which allowed containment vacuum relief valve differential pressure switches to operate in pressures that exceeded the designed operating pressure of the switches. The licensee implemented interim corrective actions to ensure operability. Specifically, the licensee increased the test frequency and adjusted the switches to reduce the effects of the deficient condition. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05106.
The performance deficiency associated with this finding involved the failure to promptly identify and correct a condition adverse to quality that could affect containment integrity. This finding was greater than minor because it affected the Configuration Control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that the containment physical design barrier protected the public from radionuclide releases caused by an event. Using the NRC Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. This finding had a crosscutting aspect in the area of human performance, associated with the decision making component, in that, licensee personnel failed to make conservative decisions related to equipment operation in accordance with design requirements (H.1(b)).
Inspection Report# : 2008007 (pdf)
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a pressure boundary valve The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, for a failure to follow procedure, when the licensee failed to complete an adequate operability evaluation for Valve SI 142A. Specifically, on August 21, 2008, the licensee failed to follow Procedure EN OP 104, "Operability Determinations," Revision 3, because personnel did not determine the leak rate solely through the required pressure boundary valve. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05077.
The failure to perform an adequate operability evaluation on safety related plant equipment in accordance with Procedure EN OP 104 is a performance deficiency. The team determined this finding was greater than minor from review of Manual Chapter 0612, Appendix E, "Examples of Minor Issues." The finding was similar to non minor finding Example 3.j in that reasonable doubt existed related to the operability of Valve SI 142A. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. The finding had a crosscutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate valve operability (P.1(c)).
 
Inspection Report# : 2008007 (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 : August 31, 2009
 
Waterford 3 3Q/2009 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Test Procedure for Safety Injection Valves S1-405A(B) Post Modification Testing Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion III due to the failure by the licensee to perform adequate post modification testing to evaluate the adequacy of design modifications made to the actuators of low pressure safety injection Isolation Valves SI-405A(B). This led to the licensee failing to identify a fundamental difference in the manner that the air operated valve actuator operated resulting in the valve popping open instead of slowly opening, creating a pressure transient that resulted in the lifting of the low temperature overpressure relief valve causing an intersystem loss-of-coolant event. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-4161.
This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both Valves SI-405A(B) inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss-of-inventory event. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. This finding had a crosscutting aspect in Human Performance, specifically the Resources aspect [H.2(a)] because the licensee failed to maintain adequate design margins. Specifically, the licensees pneumatic actuator for SI-405B could not overcome the pressure locking mechanism until twelve minutes into a fifteen minute time limit, after receiving the open demand signal. This led to the instantaneous valve disc displacement when the valve popped open causing the pressure surge, which resulted in the opening of relief valve SI-406B and subsequent loss of inventory event (Section 4OA5).
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine the Cause of a 125 Vdc Battery Failure DRAFT - Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion VXI (Corrective Actions) because the licensee failed to identify the cause for a significant condition adverse to quality. The Train B 125 Vdc battery bank failed to pass a technical specification surveillance requirement discharge test during a Spring, 2008 outage. The root cause procedure required that the licensee sequester the battery in a controlled area so that vital information related to the failure could be obtained. However, the licensee disposed of the battery instead. When questions arose concerning the specified failure cause (impurities in the battery materials), the licensee was unable to provide objective evidence to support the conclusion. Had the licensee obtained objective evidence to support their conclusion that impurities caused the battery failure, a 10 CFR Part 21 report may have been required. The licensee replaced the battery and planned to replace similar batteries in the other two trains during the next refueling outage.
The licensee entered this finding in their corrective action program as Condition Report CR WF3-2009-2846.
The finding was more than minor because, if left uncorrected, it could lead to a more significant safety concern.
Specifically, since the cause of the battery failure was not definitively found, the licensee may not have taken
 
corrective actions to prevent other battery failures. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it did not actually cause the loss of operability or functionality of another 125 Vdc battery at the time of the inspection. This finding had a crosscutting aspect in the area of Problem Identification and Resolution (Corrective Action Program Component) because the licensee failed to thoroughly evaluate the need to keep the battery prior to disposal [P.1(c)]
(Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Several Conditions Adverse to Quality DRAFT - Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to promptly correct conditions adverse to quality. The licensee had documented several conditions adverse to quality and then transferred the concerns to other condition reports. Then, the licensee closed those condition reports without addressing the concerns. Identified conditions included: 1) the Train B 125 Vac discharge test data indicated a loose battery connection but the battery was permitted to pass the test anyway; 2) the root cause determination for the failed battery was focused on the statements of one person and failed to address other information; 3) the root cause determination failed to address conflicting information; and 4) the root cause determination failed to properly address other potential causes for the inoperable battery, such as tampering. Plant personnel had failed to accurately translate the issues when transferring information from one condition report to another. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2009-1177.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. For example, the failure to include acceptance criteria in the battery discharge test (intended to identify and correct loose battery connections) could result in another inoperable 125 Vdc battery for an extended period of time. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet and determined that the finding was of very low risk significance because it did not result in another battery becoming inoperable or nonfunctional. This finding had a crosscutting aspect in the area of Human Performance (Work Practices Component) because plant personnel failed to effectively use human error prevention techniques, such as self and peer checking, when transferring concerns between condition reports [H.4(a)] (Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      May 22, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to fire protection.
The team identified a non-cited violation of License Condition 2.C.9 for the failure to identify conditions adverse to the fire protection program, as required by Procedure UNT-005-013, "Fire Protection Program," Revision 10.
Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of either bent or misaligned sprinkler head deflector plates, which were not protected as required by National Fire Protection Association 13 1976, "Standard for the Installation of Sprinkler Systems."
The failure to identify a condition adverse to fire protection was a performance deficiency. This deficiency was more than minor since, if left uncorrected, the finding would become a more significant safety concern in that the number of damaged sprinklers would continue to increase. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the team determined that the deficiency had low degradation since less
 
than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1. This finding has a cross cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions.
Inspection Report# : 2009006 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: FIN Finding Failure to Follow Commitment Tracking Procedures The inspectors identified a finding because the licensee inadvertently deleted procedural steps to recover an emergency diesel generator during a severe accident. The steps were part of a formal commitment to the NRC. The licensee had failed to follow the site commitment management program when making the procedure change and the procedure writer failed to understand the basis for the steps prior to deleting them. The licensee entered this finding in their corrective action program as Condition Reports CR WF3-2009-0193 and CR WF3-2009-1616.
The finding was more than minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, during a severe accident, operators would not have an appropriate mitigation strategy for starting an emergency diesel generator under certain severe accident conditions. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because the finding: (1) could result in a loss of functionality of an emergency diesel generator; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of Human Performance, Decision Making component [H.1(a)], because the licensee failed to use a systematic process when removing the procedural steps Inspection Report# : 2009002 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Voltage Readings Following a Single Cell Battery Charge The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V (Instructions, Procedures and Drawings) because the licensee failed to implement instructions that were intended to help troubleshoot a defective 125 Vdc battery cell. In response to the degraded cell, the licensee had established additional measures to monitor the cell following charging to ensure proper cell operation. However, the licensee did not perform the monitoring. Once identified by the inspectors, the licensee performed more frequent cell tests. The licensee subsequently replaced the faulty cell. The licensee entered this finding into their corrective action program as Condition Reports CR-WF3-2009-1088 and CR-WF3-2009-1099.
The finding was more than minor because it could have resulted in a more significant safety concern if left uncorrected. Specifically, the normal monitoring period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality [P.1(d)]
Inspection Report# : 2009002 (pdf)
 
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions
* Green. The inspectors reviewed a self revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI due to the failure by the licensee to take prompt corrective actions following the identification of an inadequate testing method used for determining the integrity of the Essential Chiller B heat exchanger tubing. Failure to take this timely action resulted in an inadvertent tube rupture and inoperability of Essential Chiller B. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5342.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because, assuming worst case degradation of both the B and AB Essential Chillers failing, the redundant A Essential Chiller would still have been available for accident mitigation. This finding had a crosscutting aspect in Problem Identification and Resolution, specifically the Corrective Action Program aspect [P.1 (d)] because the licensee failed to take appropriate corrective actions to address a degrading condition in a timely manner. Specifically, the failure to perform timely tube inspections of Essential Chiller B, following the identification of an inadequate testing methodology used for identifying Essential Chiller heat exchanger tubing degradation (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Essential Chiller AB Component Failure Due to Inadequate Procedural Guidance
* Green. The inspectors reviewed a self revealing noncited violation of Technical Specification 6.8.1.a for failure to provide documented instructions appropriate to the circumstances as recommended in Appendix A of Regulatory Guide 1.33. The failure by the licensee to provide adequate guidance for the replacement of the Essential Chiller AB compressor motor temperature sensor resulted in the reintroduction of a failure mechanism that had previously been corrected. This subsequently led to the failure of the temperature sensor wiring and inoperability of Essential Chiller AB. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5471.
This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because the redundant Essential Chillers A and B would still have been available for accident mitigation. Based on the guidance provided in Manual Chapter 0612, Appendix B, Section 1-5, Screen for Cross-Cutting Aspects, this finding did not have a crosscutting aspect because it was not considered to be reflective of current licensee performance. Specifically, the licensees failure to update the model work instructions in 2000 was a latent issue, whereby the licensee did not have a reasonable opportunity to identify the problem prior to August, 2008. In addition, the licensee has since instituted programs and processes such that the problem would not reasonably occur today (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calculations Used for Operability determination of SI-405 A(B)
Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III to address three examples of inadequate calculations involving shutdown cooling Valves SI-405A and SI-405B. The calculations were also used, in part, to support valve operability, which made an existing operability assessment invalid. First, a calculation performed by a contractor to estimate the bounding thrust requirements for pressure locking contained
 
errors and used mathematical formulas out of their intended context without applying uncertainties to account for the differences. Recent operational experience with these valves was inconsistent with the calculation's conclusions. In addition, the licensee failed to meet their quality assurance program requirements that specified that engineers perform a design verification of the calculation prior to use. Second, the licensee's calculation, that demonstrated valve actuator thrust capabilities, contained errors. Specifically, it failed to account for the friction between the actuator piston disk and walls as well as the weight of components. Third, a calculation that determined that the temperature within the valve bonnet would not heat up during small break loss of coolant accidents and faulted steam generator accidents was inadequate, in that it failed to address a faulted steam generator event, it used heat transfer calculation methods on water that were intended only for solid materials, it failed to model all components, and it failed to determine the temperatures inside the valve bonnets, which was the overriding variable of interest. The licensee entered the finding into the corrective action program as Condition Report CR-WF3-2009-00127.
This finding was more than minor because it was similar to non-minor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt concerning the operability of Valves SI-405A(B). The inspectors utilized NRC Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI-405A(B) valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high-pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution (P.1(c)) because engineers failed to thoroughly evaluate the potential for valve pressure-locking. The calculations were completed in 2008 and were indicative of current performance (Section 4OA2).
Inspection Report# : 2008005 (pdf)
Barrier Integrity Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct a condition adverse to quality.
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to promptly identify and correct a condition adverse to quality. Specifically, from March 20, 2007, through October 27, 2008, personnel failed to identify and correct a condition, which allowed containment vacuum relief valve differential pressure switches to operate in pressures that exceeded the designed operating pressure of the switches. The licensee implemented interim corrective actions to ensure operability. Specifically, the licensee increased the test frequency and adjusted the switches to reduce the effects of the deficient condition. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05106.
The performance deficiency associated with this finding involved the failure to promptly identify and correct a condition adverse to quality that could affect containment integrity. This finding was greater than minor because it affected the Configuration Control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that the containment physical design barrier protected the public from radionuclide releases caused by an event. Using the NRC Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. This finding had a crosscutting aspect in the area of human performance, associated with the decision making component, in that, licensee personnel failed to make conservative decisions related to equipment operation in accordance with design requirements (H.1(b)).
Inspection Report# : 2008007 (pdf)
 
Significance:      Oct 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a pressure boundary valve The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, for a failure to follow procedure, when the licensee failed to complete an adequate operability evaluation for Valve SI 142A. Specifically, on August 21, 2008, the licensee failed to follow Procedure EN OP 104, "Operability Determinations," Revision 3, because personnel did not determine the leak rate solely through the required pressure boundary valve. The licensee entered this deficiency into their corrective action program as Condition Report 2008 05077.
The failure to perform an adequate operability evaluation on safety related plant equipment in accordance with Procedure EN OP 104 is a performance deficiency. The team determined this finding was greater than minor from review of Manual Chapter 0612, Appendix E, "Examples of Minor Issues." The finding was similar to non minor finding Example 3.j in that reasonable doubt existed related to the operability of Valve SI 142A. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, the team determined the finding had very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment building. The finding had a crosscutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate valve operability (P.1(c)).
Inspection Report# : 2008007 (pdf)
Emergency Preparedness Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Licensee Practices Result in Protective Actions Recommendations for Areas Where Protective Action Guides Are Not Exceeded DRAFT - Green. The inspectors identified a noncited violation of 10 CFR 50.47(b)(10), for the licensees failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Specifically, the licensees guidelines for extending existing protective action recommendations into additional geographical areas of the emergency planning zone under conditions of changing wind vectors were not consistent with the guidance of EPA 400 R 92 001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents." The licensees practices resulted in unnecessary recommendations for protective actions in areas where valid dose projections show federal protective action guides are not exceeded, and may expose members of the public to unjustified risks. The licensee has entered this issue into their corrective action system as Condition Report CR-WF3-2009-03256.
This finding was more than minor because it was not similar to the examples of Manual Chapter 0612, Appendix E, and affected the emergency preparedness cornerstone objective because unnecessary protective actions may expose members of the public to an unjustified risk. The finding was associated with the emergency response organization attributes of 50.47(b) planning standards and training. This finding was of very low safety significance because it was not a risk significant planning standard functional failure or degraded function because licensee protective action recommendations would be issued in accordance with federal guidance for all areas of the emergency planning zone where Protective Action Guides are exceeded. This finding was evaluated as not having a crosscutting aspect because the finding was not indicative of current licensee performance (Section 1EP1).
Inspection Report# : 2009003 (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 : December 10, 2009
 
Waterford 3 4Q/2009 Plant Inspection Findings Initiating Events Significance:        Sep 18, 2009 Identified By: NRC Item Type: FIN Finding Failure to Incorporate Start-Up Transformer Protective Relay Design Basis into Instructions, Procedures, or Drawings.
The team identified a finding for failure to translate design basis criteria into a design basis document for the start-up transformer 3A 51G relay to support the settings listed in Calculation EC E90 012, Protective Relays Settings for Main Generator and Transformers, Revision 1. Without the design basis criteria for the 51G relay, the setpoint values could not be established. Specifically, the team determined that the relay settings listed in Calculation EC E90 012 had not been effectively implemented since the required current transformer ratio of 600/5, upon which the settings were based, was never installed. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04813.
This finding was more than minor because the failure to provide adequate relay setting coordination could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient (initiating event).
The team noted that this finding also applies to 51G relay in the B train which could challenge the single failure criterion. The team determined this finding was of very low safety significance (Green) because the issue would not prevent the safety buses from being reenergized by the emergency diesel generators. Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.10).
Inspection Report# : 2009009 (pdf)
Mitigating Systems Significance:        Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify an adverse trend in failures of time-delay relays The team identified a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform a root cause analysis and implement corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, multiple failures of Agastat E7024PB relays that were installed in or designated for safety-related applications constituted a significant condition adverse to quality. The evaluations for the individual relay failures were narrow and did not identify the adverse trend until eight relays had failed in service and seven had failed pre-installation bench tests over a two-year period. The failure of these relays would prevent auto-starting of critical equipment during a loss of offsite power, potentially creating a substantial safety hazard.
The failure of the licensee to recognize that the adverse trend in failures of Agastat E7024PB relays constituted a significant condition adverse to quality, to perform a root cause evaluation, and to initiate corrective actions to prevent recurrence is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the performance deficiency was determined to require a Phase 2 analysis because of the potential for a loss of safety system function. A Phase
 
2/Phase 3 Significance Determination was performed by an NRC Senior Reactor Analyst. Based on a bounding analysis, the analyst quantitatively determined that the actual change in core damage frequency (?CDF) due to the increased failure rate of Agastat E7024PB relays would be less than 4.0E-7/year. Therefore, this performance deficiency was determined to be of very low safety significance (Green).
This performance deficiency was determined to have a Problem Identification and Resolution cross-cutting aspect in the Corrective Action Program component because the licensee failed to periodically trend and assess information from the Corrective Action Program and other assessments in the aggregate to identify programmatic and common cause problems.
Inspection Report# : 2009010 (pdf)
Significance:      Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate extension of qualified service life of Agastat relays The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which occurred when the licensee inappropriately extended the service life of 322 safety-related Tyco/Agastat series E7000 time-delay relays without having an adequate technical basis. Specifically, the licensees engineering justification for extending the qualified life beyond the manufacturer-recommended ten years considered only degradation due to thermal aging; it failed to consider other known modes of degradation in accordance with applicable industry standards. Further, the team identified that a performance monitoring program intended to assess any increased failure rate due to this change was inappropriately canceled.
The failure of the licensee to perform a complete analysis of aging effects as required by industry standards in extending the qualified life of safety-related Agastat E7000-series relays is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of design control because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this performance deficiency was determined to be of very low safety significance (Green) because it is a design or qualification deficiency confirmed not to result in loss of operability or functionality.
Specifically, only one of the identified relay failures had occurred beyond the recommended 10-year service life; this failure did not result in the failure of multiple redundant trains of safety-related equipment . This finding was determined not to have a cross-cutting aspect because it is not indicative of current licensee performance.
Inspection Report# : 2009010 (pdf)
Significance:      Oct 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow technical specification requirements for Reactor Protective Instrumentation.
Green. The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. The licensee entered this finding in their corrective action program as condition reports CR WF3-2009-4401 and CR-WF3-2009-4407.
The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for reactor protective trips to be inadvertently removed while at power. The failure to meet the technical specifications was considered to be of very low safety significance (Green), since there was no
 
actual loss of safety function. This finding has a cross-cutting aspect in the decision-making component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to comply with technical specification 3.3.1 by declaring the log power Channel B inoperable and not placing DNBR, LPD, and reactor coolant flow channels in either bypass or trip condition (H.1.b). (Section 1R15)
Inspection Report# : 2009004 (pdf)
Significance:        Sep 24, 2009 Identified By: Licensee Item Type: VIO Violation Inoperable 125Vdc battery because electricians failed to follow work instructions White. Following a September 2, 2008 train B 125 Vdc battery failure, the licensee identified a violation of Technical Specification 6.8.1.a for the failure to follow plant procedures during corrective maintenance on the safety-related battery. Following the replacement of the entire battery bank during a 2008 refueling outage, craftsmen identified a faulty battery cell. When replacing the faulty cell, plant workers did not follow all of the specified procedural steps in the work package. The additional work resulted in a loose battery connection that rendered the entire battery bank inoperable. The licensee also failed to address an indicator of the loose connection during the battery discharge test.
The condition then went undetected for several months. The licensee entered this finding in their corrective action program as Condition Report CR WF3 2008-4179.
This finding was greater than minor because it was similar to non-minor example 4.a in NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that the failure to follow site procedures adversely affected safety related equipment. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding required a Phase 2 significance determination because it resulted in the loss of a single train of safety related equipment for greater than the technical specification allowed outage time.
Using a T/2 exposure time of 50 days, the inspectors used the Risk-Informed Inspection Notebook for Waterford Nuclear Power Plant Unit 3, Revision 2.01 and its associated Phase 2 Pre-Solved Table, and determined that a Phase 3 significance determination was necessary. A Region IV senior reactor analyst performed a preliminary Phase 3 significance determination and found that the finding was White. This preliminary Phase 3 significance determination is included as Attachment 2 to this report. This finding had a cross cutting aspect in area of Human Performance (work practices component) because maintenance personnel failed to use appropriate human error prevention techniques, such as peer checking (quality control hold points) and tracking battery components that were loosened (H.4.a). (Section 1R15).
Update: A Regulatory Conference was held for this issue on December 14, 2009. The final significance of this issue was determined to be White as described in a letter to the licensee (ML1001506600), dated January 14, 2010.
Inspection Report# : 2009008 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Account for Reduction of Flow from the Emergency Feedwater System to the Steam Generators The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee did not account for reduction of flow from the emergency feedwater system when analyzing the flow rate to the steam generators and establishing the acceptance criteria for the performance of the motor-driven emergency feedwater pumps. The factors associated with the loss of flow included the emergency diesel generator under-frequency of 0.3 Hertz allowed by technical specifications, and not accounting for accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve. The pumps had a documented analyzed margin of 55 gpm. The margin was reduced by 24 gpm due to allowed diesel under-frequency. Another reduction was attributed to the accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve.
This left the combined margin of both emergency feedwater motor-driven pumps at 6 gpm. The licensee entered this issue into the corrective action program as Condition Reports CR-WF3-2009-04731, CR-WF3-2009-04528, and CR-WF3-2009-05043, and performed an operability assessment for each of these factors.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to
 
ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. This finding is of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation. Some margin in total flow still remained to compensate for the reduced pump performance if operated at the reduced-frequency.
The inspectors determined that the finding has a cross cutting aspect in the area of Problem Identification and Resolution, associated with Operating Experience. The licensee had received NRC Information Notice 2008-02, which specifically identified the diesel under-frequency as a potential problem for ac motor-operated pumps, and test acceptance criteria concerns which would have ensured the capability of the equipment to perform its function under the most limiting conditions. The licensee failed to identify the applicability of these potential problems to the emergency feedwater motor-operated pumps and take proper actions [P.2(a).] (Section 1R21.b.1.1).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Proper Design Control Measures to Assure Adequate Design and to Properly Translate the Design into Test Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with three examples.
Example 1: The licensee did not use the correct size emergency feedwater system suction piping in calculation MNQ10-12 Net Positive Suction Head Available for Emergency Feedwater Pumps. The motor-driven pump suction piping is 4 inches in diameter but the licensee nonconservatively used 6-inch piping in the calculations. The licensee has entered this issue into their corrective action program as Condition Report CR-WF3-2009-04729 and performed an operability assessment for the issue.
Example 2: Calculation ECM91-001, Revision 3, Emergency Diesel Generator Fuel Oil Transfer Pump Recirculation and Discharge Flow, arbitrarily assumed that the suction strainer of the fuel oil transfer pump would only be 10 percent clogged. The licensee could not justify the 10 percent clogging assumption or find any justification for selecting the 10 percent value. Also, there is no discussion or any physical comparison to ensure that the mesh of the installed Leslie strainer was the same as that of the Hayward strainer identified in an attachment to the calculation. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04812 and performed an operability assessment for the issue.
Example 3: Calculation EC-I01-003, Revision 0, IST Instrumentation Uncertainties, determines the adequacy of permanent plant instrumentation for inservice testing use. The calculation determined that some specific instruments shall not be used for inservice testing applications. Contrary to the calculation requirements, procedure OP 903 014, used for the inservice testing comprehensive test of the emergency feedwater pumps, specified that the forbidden flow instruments shall be used for verification of emergency feedwater system flow rate. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04811. These findings are more than minor because they affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance deficiencies would have the potential to lead to more significant safety concerns. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, April 20 23, 2009, that included the emergency feedwater turbine-driven pump and the emergency diesel
 
generator fuel oil transfer pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify any of these three issues [P.3.(a).] (Section 1R21.b.1.6).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to have an Operating Procedure for Executing an Evolution Credited in the UFSAR and in an Request for a License Amendment The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings pertaining to the emergency diesel generator fuel oil transfer pump. Criterion V states, in part, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Specifically, the licensee did not have operating procedures for accomplishing the transfer of fuel oil from one storage tank to the opposite train feed tank (day tank) using the opposite train fuel oil transfer pump, as designated in the USAR Table 9.5-2, Failure Mode and Effects Analysis. Also, License Amendment Number 157 (TAC Number MA4940) was granted, in part, for having the capability to transfer fuel oil from one storage tank to the opposite train feed tank using the opposite transfer pump. The licensee specified this capability as part of the justification for having an insufficiently sized fuel oil storage tank. Moreover, the Safety Evaluation Report associated with License Amendment Number 157 specifically referred to this capability at Waterford 3, and specified that procedures were available for accomplishing the transfer of fuel oil. The licensee has entered this finding in their corrective action program as Condition Report CR-WF3-2009-04950, and performed an operability assessment for the issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events.
Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because the failure to have an operating procedure did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.7).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Analyze the Effect of Acceptable Reverse Flow through Emergency Feedwater Check Valves The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to analyze the effects of the acceptable back leakage of 25 gpm from the emergency feedwater pump discharge check valves on the integrity of the emergency feedwater pumps and the integrity of its suction piping. The acceptable back leakage could possibly cause the pump to reverse rotate, and provide a path for high pressure fluid to go through the pump and pressurize low pressure suction piping. The licensee has entered this item in their corrective action program as Condition Report CR WF3 2009 04528 and performed an operability assessment for this issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if
 
uncorrected. This finding was determined to be of very low safety significance (Green) because this design issue did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, on April 20-23, 2009, that included the emergency feedwater AB turbine-driven pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify the impact of reverse flow on the integrity of the pump and its suction piping [P.3.(a)] (Section 1R21.b.1.8).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify or Check the adequacy of Design Changes for the Emergency Diesel Generator Protective Relay IGVC-51V The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. The calculation EE2 14 3 Diesel Generator Overcurrent Protection, Revision 1, does not document sufficient design bases for the setting of the IGCV 51 overcurrent with voltage control relays for the emergency diesel generators.
Specifically, the licensee failed to perform an adequate evaluation of new setpoint values identified in Engineering Report ER W3 99 0174 00 00, which provided the bases for relay tap setpoint changes for emergency diesel generator overcurrent protection while the diesel was in test mode. The primary purpose of the IGCV-51V relays was to protect the emergency diesel generator against external faults and prevent the output breaker from closing following a breaker trip associated with a fault. If the faulted bus had been isolated by the operation of the under-voltage relays instead of the IGCV 51 relays, the emergency diesel generator output breaker would be allowed to electrically reclose onto this faulted bus and potentially damage the emergency diesel generator and the associated switchgear. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04780.
The failure to have sufficient design bases for the emergency diesel generator overcurrent protection IGCV 51V relays without an adequate verification of the setpoint modification for the IGCV 51V relay, Voltage Controlled, Time-Overcurrent Relay, for emergency diesel generator overcurrent protection while the diesel was in test mode, was a performance deficiency. Specifically, failure to verify the adequacy of a design modification for the IGCV 51V relay could result in reduced reliability of the emergency diesel generators. The finding was determined to be greater than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences.
Using Manual Chapter 0609.04, the finding was determined to have a very low safety significance (Green) because the failure did not result in loss of operability or functionality and because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.12).
Inspection Report# : 2009009 (pdf)
Significance:        Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine the Cause of a 125 Vdc Battery Failure Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion VXI (Corrective Actions) because the licensee failed to identify the cause for a significant condition adverse to quality. The Train B 125 Vdc battery bank failed to pass a technical specification surveillance requirement discharge test during a Spring, 2008 outage. The root cause procedure required that the licensee sequester the battery in a controlled area so that vital information related to the failure could be obtained. However, the licensee disposed of the battery instead. When questions arose concerning the specified failure cause (impurities in the battery materials), the licensee was unable to provide objective evidence to support the conclusion. Had the licensee obtained objective evidence to support their conclusion that impurities caused the battery failure, a 10 CFR Part 21 report may have been required. The licensee
 
replaced the battery and planned to replace similar batteries in the other two trains during the next refueling outage.
The licensee entered this finding in their corrective action program as Condition Report CR WF3-2009-2846.
The finding was more than minor because, if left uncorrected, it could lead to a more significant safety concern.
Specifically, since the cause of the battery failure was not definitively found, the licensee may not have taken corrective actions to prevent other battery failures. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it did not actually cause the loss of operability or functionality of another 125 Vdc battery at the time of the inspection. This finding had a crosscutting aspect in the area of Problem Identification and Resolution (Corrective Action Program Component) because the licensee failed to thoroughly evaluate the need to keep the battery prior to disposal [P.1(c)]
(Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Several Conditions Adverse to Quality Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to promptly correct conditions adverse to quality. The licensee had documented several conditions adverse to quality and then transferred the concerns to other condition reports. Then, the licensee closed those condition reports without addressing the concerns. Identified conditions included: 1) the Train B 125 Vac discharge test data indicated a loose battery connection but the battery was permitted to pass the test anyway; 2) the root cause determination for the failed battery was focused on the statements of one person and failed to address other information; 3) the root cause determination failed to address conflicting information; and 4) the root cause determination failed to properly address other potential causes for the inoperable battery, such as tampering. Plant personnel had failed to accurately translate the issues when transferring information from one condition report to another. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2009-1177.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. For example, the failure to include acceptance criteria in the battery discharge test (intended to identify and correct loose battery connections) could result in another inoperable 125 Vdc battery for an extended period of time. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet and determined that the finding was of very low risk significance because it did not result in another battery becoming inoperable or nonfunctional. This finding had a crosscutting aspect in the area of Human Performance (Work Practices Component) because plant personnel failed to effectively use human error prevention techniques, such as self and peer checking, when transferring concerns between condition reports [H.4(a)] (Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      May 22, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to fire protection.
The team identified a non-cited violation of License Condition 2.C.9 for the failure to identify conditions adverse to the fire protection program, as required by Procedure UNT-005-013, "Fire Protection Program," Revision 10.
Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of either bent or misaligned sprinkler head deflector plates, which were not protected as required by National Fire Protection Association 13 1976, "Standard for the Installation of Sprinkler Systems."
The failure to identify a condition adverse to fire protection was a performance deficiency. This deficiency was more than minor since, if left uncorrected, the finding would become a more significant safety concern in that the number of
 
damaged sprinklers would continue to increase. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the team determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1. This finding has a cross cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions [H.2(c)].
Inspection Report# : 2009006 (pdf)
Significance: N/A May 22, 2009 Identified By: NRC Item Type: FIN Finding Failure to provide area wide sprinkler coverage as required in an Appendix R, Section III.G.2.c fire area.
The team identified a violation of License Condition 2.C.9 for failure to protect post fire safe shutdown equipment against fire damage, as required by 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, in Fire Area RAB 39 the licensee failed to provide area wide sprinkler coverage that complied with the requirements in National Fire Protection Association 13 1976. As required in Appendix R, Section III.G.2.c, redundant trains within the same fire area must be protected with detection and an automatic fire suppression system when redundant post fire safe shutdown equipment is protected with 1 hour fire barriers. The team determined this violation met the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" conditions for receiving enforcement discretion (EA 09 171).
Failure to provide area wide sprinkler coverage in accordance with National Fire Protection Association 13 1976 for a fire area with 1 hour fire barriers was a performance deficiency. The team determined that this finding was more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this violation meets the discretion criteria of the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" for a noncompliance identified during the transition to National Fire Protection Association 805, the team determined that discretion to take no enforcement action is appropriate at this time, as described in the Enforcement Policy. The team reviewed the risk assessment for the fire area and determined that the licensee demonstrated that the risk was less than high safety significance (Red). Specifically, the team determined that the fixed and transient fire sources would not generate sufficient heat to cause fire damage that rendered the systems incapable of performing their safety function.
Inspection Report# : 2009006 (pdf)
Significance: N/A May 22, 2009 Identified By: NRC Item Type: FIN Finding Failure to ensure post-fire safe shutdown valves could be operated.
The team identified a violation of License Condition 2.C.9 related to the capability to complete required manual actions, following a control room fire, because of potential fire damage to some motor operated valves. Specifically, the licensee failed to evaluate the susceptibility of fire damaging circuits in motor operated valves that needed to be manually operated for post fire safe shutdown. The licensee did not recognize that the circuits could cause the valves to become stuck. The team determined licensee personnel would not be able to reposition motor operated valves as specified in plant procedures. The team determined this violation met the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" conditions for receiving enforcement discretion (EA 09 171).
The failure to ensure that safe shutdown equipment could be operated as required during control room fire events was a performance deficiency. The team determined that this finding was more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this violation meets the discretion criteria of the "Interim Enforcement Policy Regarding
 
Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" for a noncompliance identified during the transition to National Fire Protection Association 805, the team determined that discretion to take no enforcement action is appropriate at this time, as described in the Enforcement Policy.
Inspection Report# : 2009006 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: FIN Finding Failure to Follow Commitment Tracking Procedures The inspectors identified a finding because the licensee inadvertently deleted procedural steps to recover an emergency diesel generator during a severe accident. The steps were part of a formal commitment to the NRC. The licensee had failed to follow the site commitment management program when making the procedure change and the procedure writer failed to understand the basis for the steps prior to deleting them. The licensee entered this finding in their corrective action program as Condition Reports CR WF3-2009-0193 and CR WF3-2009-1616.
The finding was more than minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, during a severe accident, operators would not have an appropriate mitigation strategy for starting an emergency diesel generator under certain severe accident conditions. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because the finding: (1) could result in a loss of functionality of an emergency diesel generator; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of Human Performance, Decision Making component [H.1(a)], because the licensee failed to use a systematic process when removing the procedural steps Inspection Report# : 2009002 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Voltage Readings Following a Single Cell Battery Charge The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V (Instructions, Procedures and Drawings) because the licensee failed to implement instructions that were intended to help troubleshoot a defective 125 Vdc battery cell. In response to the degraded cell, the licensee had established additional measures to monitor the cell following charging to ensure proper cell operation. However, the licensee did not perform the monitoring. Once identified by the inspectors, the licensee performed more frequent cell tests. The licensee subsequently replaced the faulty cell. The licensee entered this finding into their corrective action program as Condition Reports CR-WF3-2009-1088 and CR-WF3-2009-1099.
The finding was more than minor because it could have resulted in a more significant safety concern if left uncorrected. Specifically, the normal monitoring period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality [P.1(d)]
Inspection Report# : 2009002 (pdf)
 
Barrier Integrity Emergency Preparedness Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Licensee Practices Result in Protective Actions Recommendations for Areas Where Protective Action Guides Are Not Exceeded Green. The inspectors identified a noncited violation of 10 CFR 50.47(b)(10), for the licensees failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Specifically, the licensees guidelines for extending existing protective action recommendations into additional geographical areas of the emergency planning zone under conditions of changing wind vectors were not consistent with the guidance of EPA 400 R 92 001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents." The licensees practices resulted in unnecessary recommendations for protective actions in areas where valid dose projections show federal protective action guides are not exceeded, and may expose members of the public to unjustified risks. The licensee has entered this issue into their corrective action system as Condition Report CR-WF3-2009-03256.
This finding was more than minor because it was not similar to the examples of Manual Chapter 0612, Appendix E, and affected the emergency preparedness cornerstone objective because unnecessary protective actions may expose members of the public to an unjustified risk. The finding was associated with the emergency response organization attributes of 50.47(b) planning standards and training. This finding was of very low safety significance because it was not a risk significant planning standard functional failure or degraded function because licensee protective action recommendations would be issued in accordance with federal guidance for all areas of the emergency planning zone where Protective Action Guides are exceeded. This finding was evaluated as not having a crosscutting aspect because the finding was not indicative of current licensee performance (Section 1EP1).
Inspection Report# : 2009003 (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 : March 01, 2010
 
Waterford 3 1Q/2010 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Transient Combustibles (Section 1R05)
* Green. The inspectors identified five examples of a green noncited violation of Waterford 3 Steam Electric Station license condition 2.C.9 for the failure to perform a transient combustible evaluation prior to introducing transient combustibles into procedurally controlled vital plant areas. Specifically, procedures limit the amount of transient combustibles that may be introduced into the control room ventilation equipment room, the component cooling water Train B heat exchanger room, and the main steam isolation valve Train B roof area. Any amounts greater than the preset procedural limits require a transient combustible evaluation to be performed. In all five cases, this evaluation was not performed prior to introduction of the transient combustibles. This violation has been entered into the licensees corrective action program as condition reports CR WF3 2010 0482, CR-WF3-2010-0598, and CR-WF3-2009-4035.
The performance deficiencies associated with this violation were the failure to comply with Waterford 3 Steam Electric Station license condition 2.C.9. Specifically, the procedural requirements to perform a transient combustible evaluation prior to introducing the transient combustibles into designated fire zones were not performed. Since several of the previously described fire zones fail to meet Appendix R train separation requirements, use of Inspection Manual Chapter 0612, Appendix E to screen for minor examples is not appropriate. This condition is greater than minor because, if left uncorrected, it would become a more significant safety concern, since continued introduction of unevaluated transient combustible loading into controlled areas could significantly reduce the ability to achieve a safe shutdown condition, in the event of a fire in that controlled area. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, to assess the safety significance. Since the severity of the observed deficiencies was assigned a low degradation rating, it was determined to be of very low risk significance. This finding had a crosscutting aspect in the area of human performance associated with the work practices component in that the licensee failed to utilize appropriate human error prevention techniques by (1) discussing transient combustible controls and expectations during pre-job briefs and (2) effectively utilizing human performance barriers, such as posted door signs.
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: NRC Item Type: FIN Finding Failure to Implement Procedure to Restore the 5A Feedwater Heater Level Switch (Section 1R19)
* Green. A self-revealing finding was identified for the failure to implement work order instructions relied upon to open the upper isolation valve for the 5A feedwater heater level switch following functional checks of the level switch. The failure to implement work order instructions necessary to return the 5A feedwater heater level switch to service resulted in a subsequent automatic isolation of a low pressure feedwater heater string on December 7, 2009 which required a plant down power to 72% in order to comply with limiting plant conditions specified by station procedure OP-003-034, Feed Heater Vents and Drains. This finding has been entered into the licensees CAP as CR-WF3-2009-7420.
The failure to implement step 4.4 of work order 00180716-01 on October 30, 2009 to open the upper isolation valve (FHD-703A) for the 5A feedwater heater level switch (FHDILS1553A) was a performance deficiency. This finding was greater than minor because it resulted in the automatic isolation of a low pressure feedwater heater string requiring a plant down power and; therefore, adversely affected the initiating events cornerstone objective to limit the likelihood of those plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. This finding was assessed using the initiating events cornerstone column of the Phase 1
 
screening worksheet of the SDP and determined to be of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. This finding was assigned a cross-cutting aspect in the work practices component of the human performance area because the licensee failed to effectively communicate and implement the use of concurrent verification of valve position as a human error prevention technique to ensure that personnel work practices supported human performance. The procedure the licensee failed to implement was not safety related, therefore, the finding did not result in a violation of regulatory requirements.
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Update Drawings after Design Change Green. A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for the licensees failure to prescribe an activity affecting quality by documented instructions, procedures, or drawings appropriate to the circumstance. Specifically, for all reactor coolant pump heat exchanger to pump cover bolted connection gasket replacements between the refueling outage of 1986 (RF-1) and the refueling outage of 2009 (RF-16), the licensee prescribed the wrong gasket material, gasket size, and fastener preload because they had failed to incorporate a design change implemented during RF-1 into their instructions, procedures, or drawings. Station modification package SMP-1427, an engineering change implemented during RF-1 in response to industry operating experience, called for a thicker gasket, different gasket material, and an increased bolt preload in order to increase gasket compression and reduce the probability of leakage. As a consequence of failing to incorporate SMP-1427 changes into procedures, all heat exchanger gasket replacements since RF-1, four gasket replacements in total, have utilized thinner gaskets with less than the vendor recommended compression. The licensee documented this condition in CR-WF3-2009-5501.
The licensees failure to prescribe appropriate gasket replacement requirements is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available.
This finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee did not institutionalize operating experience through changes to the station procedures
[P.2(b)].
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Reactor Coolant Pump Vapor Seal Leakage Green. A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee did not promptly correct reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on the component cooling water heat exchanger and cover areas of three reactor coolant pumps. Corrective actions for this condition were implemented during refuel outage 15, but these corrective actions failed to correct the condition and the vapor seal leakage continued through operating cycle 16. This resulted in some additional boric acid corrosion and degradation to reactor coolant pump covers and carbon steel component cooling water flanges. The licensee implemented a design modification to correct the condition and documented the condition in CR-WF3-2009-5501.
The licensees failure to promptly correct a condition adverse to quality is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment was still available. This finding had a crosscutting aspect in area of human performance associated with work control in that the licensee did
 
not effectively plan for the resources necessary to implement the post maintenance testing associated with the corrective actions implemented during refuel outage 15, and therefore failed to discover that those corrective actions were inadequate to correct the condition[H.3(a)].
Inspection Report# : 2009005 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: FIN Finding Failure to Incorporate Start-Up Transformer Protective Relay Design Basis into Instructions, Procedures, or Drawings.
The team identified a finding for failure to translate design basis criteria into a design basis document for the start-up transformer 3A 51G relay to support the settings listed in Calculation EC E90 012, Protective Relays Settings for Main Generator and Transformers, Revision 1. Without the design basis criteria for the 51G relay, the setpoint values could not be established. Specifically, the team determined that the relay settings listed in Calculation EC E90 012 had not been effectively implemented since the required current transformer ratio of 600/5, upon which the settings were based, was never installed. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04813.
This finding was more than minor because the failure to provide adequate relay setting coordination could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient (initiating event).
The team noted that this finding also applies to 51G relay in the B train which could challenge the single failure criterion. The team determined this finding was of very low safety significance (Green) because the issue would not prevent the safety buses from being reenergized by the emergency diesel generators. Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.10).
Inspection Report# : 2009009 (pdf)
Mitigating Systems Significance:        Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify an adverse trend in failures of time-delay relays The team identified a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform a root cause analysis and implement corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, multiple failures of Agastat E7024PB relays that were installed in or designated for safety-related applications constituted a significant condition adverse to quality. The evaluations for the individual relay failures were narrow and did not identify the adverse trend until eight relays had failed in service and seven had failed pre-installation bench tests over a two-year period. The failure of these relays would prevent auto-starting of critical equipment during a loss of offsite power, potentially creating a substantial safety hazard.
The failure of the licensee to recognize that the adverse trend in failures of Agastat E7024PB relays constituted a significant condition adverse to quality, to perform a root cause evaluation, and to initiate corrective actions to prevent recurrence is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the performance deficiency was determined to require a Phase 2 analysis because of the potential for a loss of safety system function. A Phase 2/Phase 3 Significance Determination was performed by an NRC Senior Reactor Analyst. Based on a bounding analysis, the analyst quantitatively determined that the actual change in core damage frequency (?CDF) due to the increased failure rate of Agastat E7024PB relays would be less than 4.0E-7/year. Therefore, this performance
 
deficiency was determined to be of very low safety significance (Green).
This performance deficiency was determined to have a Problem Identification and Resolution cross-cutting aspect in the Corrective Action Program component because the licensee failed to periodically trend and assess information from the Corrective Action Program and other assessments in the aggregate to identify programmatic and common cause problems.
Inspection Report# : 2009010 (pdf)
Significance:      Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate extension of qualified service life of Agastat relays The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which occurred when the licensee inappropriately extended the service life of 322 safety-related Tyco/Agastat series E7000 time-delay relays without having an adequate technical basis. Specifically, the licensees engineering justification for extending the qualified life beyond the manufacturer-recommended ten years considered only degradation due to thermal aging; it failed to consider other known modes of degradation in accordance with applicable industry standards. Further, the team identified that a performance monitoring program intended to assess any increased failure rate due to this change was inappropriately canceled.
The failure of the licensee to perform a complete analysis of aging effects as required by industry standards in extending the qualified life of safety-related Agastat E7000-series relays is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of design control because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this performance deficiency was determined to be of very low safety significance (Green) because it is a design or qualification deficiency confirmed not to result in loss of operability or functionality.
Specifically, only one of the identified relay failures had occurred beyond the recommended 10-year service life; this failure did not result in the failure of multiple redundant trains of safety-related equipment . This finding was determined not to have a cross-cutting aspect because it is not indicative of current licensee performance.
Inspection Report# : 2009010 (pdf)
Significance:      Oct 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow technical specification requirements for Reactor Protective Instrumentation.
Green. The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. The licensee entered this finding in their corrective action program as condition reports CR WF3-2009-4401 and CR-WF3-2009-4407.
The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for reactor protective trips to be inadvertently removed while at power. The failure to meet the technical specifications was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to comply with technical specification 3.3.1 by declaring the log power Channel B inoperable
 
and not placing DNBR, LPD, and reactor coolant flow channels in either bypass or trip condition (H.1.b). (Section 1R15)
Inspection Report# : 2009004 (pdf)
Significance:        Sep 24, 2009 Identified By: Licensee Item Type: VIO Violation Inoperable 125Vdc battery because electricians failed to follow work instructions White. Following a September 2, 2008 train B 125 Vdc battery failure, the licensee identified a violation of Technical Specification 6.8.1.a for the failure to follow plant procedures during corrective maintenance on the safety-related battery. Following the replacement of the entire battery bank during a 2008 refueling outage, craftsmen identified a faulty battery cell. When replacing the faulty cell, plant workers did not follow all of the specified procedural steps in the work package. The additional work resulted in a loose battery connection that rendered the entire battery bank inoperable. The licensee also failed to address an indicator of the loose connection during the battery discharge test.
The condition then went undetected for several months. The licensee entered this finding in their corrective action program as Condition Report CR WF3 2008-4179.
This finding was greater than minor because it was similar to non-minor example 4.a in NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that the failure to follow site procedures adversely affected safety related equipment. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding required a Phase 2 significance determination because it resulted in the loss of a single train of safety related equipment for greater than the technical specification allowed outage time.
Using a T/2 exposure time of 50 days, the inspectors used the Risk-Informed Inspection Notebook for Waterford Nuclear Power Plant Unit 3, Revision 2.01 and its associated Phase 2 Pre-Solved Table, and determined that a Phase 3 significance determination was necessary. A Region IV senior reactor analyst performed a preliminary Phase 3 significance determination and found that the finding was White. This preliminary Phase 3 significance determination is included as Attachment 2 to this report. This finding had a cross cutting aspect in area of Human Performance (work practices component) because maintenance personnel failed to use appropriate human error prevention techniques, such as peer checking (quality control hold points) and tracking battery components that were loosened (H.4.a). (Section 1R15).
Update: A Regulatory Conference was held for this issue on December 14, 2009. The final significance of this issue was determined to be White as described in a letter to the licensee (ML1001506600), dated January 14, 2010.
Inspection Report# : 2009008 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Account for Reduction of Flow from the Emergency Feedwater System to the Steam Generators The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee did not account for reduction of flow from the emergency feedwater system when analyzing the flow rate to the steam generators and establishing the acceptance criteria for the performance of the motor-driven emergency feedwater pumps. The factors associated with the loss of flow included the emergency diesel generator under-frequency of 0.3 Hertz allowed by technical specifications, and not accounting for accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve. The pumps had a documented analyzed margin of 55 gpm. The margin was reduced by 24 gpm due to allowed diesel under-frequency. Another reduction was attributed to the accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve.
This left the combined margin of both emergency feedwater motor-driven pumps at 6 gpm. The licensee entered this issue into the corrective action program as Condition Reports CR-WF3-2009-04731, CR-WF3-2009-04528, and CR-WF3-2009-05043, and performed an operability assessment for each of these factors.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the
 
operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. This finding is of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation. Some margin in total flow still remained to compensate for the reduced pump performance if operated at the reduced-frequency.
The inspectors determined that the finding has a cross cutting aspect in the area of Problem Identification and Resolution, associated with Operating Experience. The licensee had received NRC Information Notice 2008-02, which specifically identified the diesel under-frequency as a potential problem for ac motor-operated pumps, and test acceptance criteria concerns which would have ensured the capability of the equipment to perform its function under the most limiting conditions. The licensee failed to identify the applicability of these potential problems to the emergency feedwater motor-operated pumps and take proper actions [P.2(a).] (Section 1R21.b.1.1).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Proper Design Control Measures to Assure Adequate Design and to Properly Translate the Design into Test Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with three examples.
Example 1: The licensee did not use the correct size emergency feedwater system suction piping in calculation MNQ10-12 Net Positive Suction Head Available for Emergency Feedwater Pumps. The motor-driven pump suction piping is 4 inches in diameter but the licensee nonconservatively used 6-inch piping in the calculations. The licensee has entered this issue into their corrective action program as Condition Report CR-WF3-2009-04729 and performed an operability assessment for the issue.
Example 2: Calculation ECM91-001, Revision 3, Emergency Diesel Generator Fuel Oil Transfer Pump Recirculation and Discharge Flow, arbitrarily assumed that the suction strainer of the fuel oil transfer pump would only be 10 percent clogged. The licensee could not justify the 10 percent clogging assumption or find any justification for selecting the 10 percent value. Also, there is no discussion or any physical comparison to ensure that the mesh of the installed Leslie strainer was the same as that of the Hayward strainer identified in an attachment to the calculation. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04812 and performed an operability assessment for the issue.
Example 3: Calculation EC-I01-003, Revision 0, IST Instrumentation Uncertainties, determines the adequacy of permanent plant instrumentation for inservice testing use. The calculation determined that some specific instruments shall not be used for inservice testing applications. Contrary to the calculation requirements, procedure OP 903 014, used for the inservice testing comprehensive test of the emergency feedwater pumps, specified that the forbidden flow instruments shall be used for verification of emergency feedwater system flow rate. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04811. These findings are more than minor because they affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance deficiencies would have the potential to lead to more significant safety concerns. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, April 20 23, 2009, that included the emergency feedwater turbine-driven pump and the emergency diesel generator fuel oil transfer pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify any of these three issues [P.3.(a).] (Section 1R21.b.1.6).
 
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to have an Operating Procedure for Executing an Evolution Credited in the UFSAR and in an Request for a License Amendment The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings pertaining to the emergency diesel generator fuel oil transfer pump. Criterion V states, in part, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Specifically, the licensee did not have operating procedures for accomplishing the transfer of fuel oil from one storage tank to the opposite train feed tank (day tank) using the opposite train fuel oil transfer pump, as designated in the USAR Table 9.5-2, Failure Mode and Effects Analysis. Also, License Amendment Number 157 (TAC Number MA4940) was granted, in part, for having the capability to transfer fuel oil from one storage tank to the opposite train feed tank using the opposite transfer pump. The licensee specified this capability as part of the justification for having an insufficiently sized fuel oil storage tank. Moreover, the Safety Evaluation Report associated with License Amendment Number 157 specifically referred to this capability at Waterford 3, and specified that procedures were available for accomplishing the transfer of fuel oil. The licensee has entered this finding in their corrective action program as Condition Report CR-WF3-2009-04950, and performed an operability assessment for the issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events.
Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because the failure to have an operating procedure did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.7).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Analyze the Effect of Acceptable Reverse Flow through Emergency Feedwater Check Valves The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to analyze the effects of the acceptable back leakage of 25 gpm from the emergency feedwater pump discharge check valves on the integrity of the emergency feedwater pumps and the integrity of its suction piping. The acceptable back leakage could possibly cause the pump to reverse rotate, and provide a path for high pressure fluid to go through the pump and pressurize low pressure suction piping. The licensee has entered this item in their corrective action program as Condition Report CR WF3 2009 04528 and performed an operability assessment for this issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. This finding was determined to be of very low safety significance (Green) because this design issue did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
 
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, on April 20-23, 2009, that included the emergency feedwater AB turbine-driven pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify the impact of reverse flow on the integrity of the pump and its suction piping [P.3.(a)] (Section 1R21.b.1.8).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify or Check the adequacy of Design Changes for the Emergency Diesel Generator Protective Relay IGVC-51V The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. The calculation EE2 14 3 Diesel Generator Overcurrent Protection, Revision 1, does not document sufficient design bases for the setting of the IGCV 51 overcurrent with voltage control relays for the emergency diesel generators.
Specifically, the licensee failed to perform an adequate evaluation of new setpoint values identified in Engineering Report ER W3 99 0174 00 00, which provided the bases for relay tap setpoint changes for emergency diesel generator overcurrent protection while the diesel was in test mode. The primary purpose of the IGCV-51V relays was to protect the emergency diesel generator against external faults and prevent the output breaker from closing following a breaker trip associated with a fault. If the faulted bus had been isolated by the operation of the under-voltage relays instead of the IGCV 51 relays, the emergency diesel generator output breaker would be allowed to electrically reclose onto this faulted bus and potentially damage the emergency diesel generator and the associated switchgear. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04780.
The failure to have sufficient design bases for the emergency diesel generator overcurrent protection IGCV 51V relays without an adequate verification of the setpoint modification for the IGCV 51V relay, Voltage Controlled, Time-Overcurrent Relay, for emergency diesel generator overcurrent protection while the diesel was in test mode, was a performance deficiency. Specifically, failure to verify the adequacy of a design modification for the IGCV 51V relay could result in reduced reliability of the emergency diesel generators. The finding was determined to be greater than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences.
Using Manual Chapter 0609.04, the finding was determined to have a very low safety significance (Green) because the failure did not result in loss of operability or functionality and because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.12).
Inspection Report# : 2009009 (pdf)
Significance:        Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine the Cause of a 125 Vdc Battery Failure Green. The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion VXI (Corrective Actions) because the licensee failed to identify the cause for a significant condition adverse to quality. The Train B 125 Vdc battery bank failed to pass a technical specification surveillance requirement discharge test during a Spring, 2008 outage. The root cause procedure required that the licensee sequester the battery in a controlled area so that vital information related to the failure could be obtained. However, the licensee disposed of the battery instead. When questions arose concerning the specified failure cause (impurities in the battery materials), the licensee was unable to provide objective evidence to support the conclusion. Had the licensee obtained objective evidence to support their conclusion that impurities caused the battery failure, a 10 CFR Part 21 report may have been required. The licensee replaced the battery and planned to replace similar batteries in the other two trains during the next refueling outage.
The licensee entered this finding in their corrective action program as Condition Report CR WF3-2009-2846.
The finding was more than minor because, if left uncorrected, it could lead to a more significant safety concern.
 
Specifically, since the cause of the battery failure was not definitively found, the licensee may not have taken corrective actions to prevent other battery failures. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it did not actually cause the loss of operability or functionality of another 125 Vdc battery at the time of the inspection. This finding had a crosscutting aspect in the area of Problem Identification and Resolution (Corrective Action Program Component) because the licensee failed to thoroughly evaluate the need to keep the battery prior to disposal [P.1(c)]
(Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Several Conditions Adverse to Quality Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to promptly correct conditions adverse to quality. The licensee had documented several conditions adverse to quality and then transferred the concerns to other condition reports. Then, the licensee closed those condition reports without addressing the concerns. Identified conditions included: 1) the Train B 125 Vac discharge test data indicated a loose battery connection but the battery was permitted to pass the test anyway; 2) the root cause determination for the failed battery was focused on the statements of one person and failed to address other information; 3) the root cause determination failed to address conflicting information; and 4) the root cause determination failed to properly address other potential causes for the inoperable battery, such as tampering. Plant personnel had failed to accurately translate the issues when transferring information from one condition report to another. The licensee entered this finding into their corrective action program as Condition Report CR-WF3-2009-1177.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. For example, the failure to include acceptance criteria in the battery discharge test (intended to identify and correct loose battery connections) could result in another inoperable 125 Vdc battery for an extended period of time. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet and determined that the finding was of very low risk significance because it did not result in another battery becoming inoperable or nonfunctional. This finding had a crosscutting aspect in the area of Human Performance (Work Practices Component) because plant personnel failed to effectively use human error prevention techniques, such as self and peer checking, when transferring concerns between condition reports [H.4(a)] (Section 4OA2).
Inspection Report# : 2009003 (pdf)
Significance:      May 22, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to fire protection.
The team identified a non-cited violation of License Condition 2.C.9 for the failure to identify conditions adverse to the fire protection program, as required by Procedure UNT-005-013, "Fire Protection Program," Revision 10.
Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of either bent or misaligned sprinkler head deflector plates, which were not protected as required by National Fire Protection Association 13 1976, "Standard for the Installation of Sprinkler Systems."
The failure to identify a condition adverse to fire protection was a performance deficiency. This deficiency was more than minor since, if left uncorrected, the finding would become a more significant safety concern in that the number of damaged sprinklers would continue to increase. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Attachment 2, "Degradation Rating Guidance Specific to
 
Various Fire Protection Program Elements," the team determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1. This finding has a cross cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions [H.2(c)].
Inspection Report# : 2009006 (pdf)
Significance: N/A May 22, 2009 Identified By: NRC Item Type: FIN Finding Failure to provide area wide sprinkler coverage as required in an Appendix R, Section III.G.2.c fire area.
The team identified a violation of License Condition 2.C.9 for failure to protect post fire safe shutdown equipment against fire damage, as required by 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, in Fire Area RAB 39 the licensee failed to provide area wide sprinkler coverage that complied with the requirements in National Fire Protection Association 13 1976. As required in Appendix R, Section III.G.2.c, redundant trains within the same fire area must be protected with detection and an automatic fire suppression system when redundant post fire safe shutdown equipment is protected with 1 hour fire barriers. The team determined this violation met the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" conditions for receiving enforcement discretion (EA 09 171).
Failure to provide area wide sprinkler coverage in accordance with National Fire Protection Association 13 1976 for a fire area with 1 hour fire barriers was a performance deficiency. The team determined that this finding was more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this violation meets the discretion criteria of the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" for a noncompliance identified during the transition to National Fire Protection Association 805, the team determined that discretion to take no enforcement action is appropriate at this time, as described in the Enforcement Policy. The team reviewed the risk assessment for the fire area and determined that the licensee demonstrated that the risk was less than high safety significance (Red). Specifically, the team determined that the fixed and transient fire sources would not generate sufficient heat to cause fire damage that rendered the systems incapable of performing their safety function.
Inspection Report# : 2009006 (pdf)
Significance: N/A May 22, 2009 Identified By: NRC Item Type: FIN Finding Failure to ensure post-fire safe shutdown valves could be operated.
The team identified a violation of License Condition 2.C.9 related to the capability to complete required manual actions, following a control room fire, because of potential fire damage to some motor operated valves. Specifically, the licensee failed to evaluate the susceptibility of fire damaging circuits in motor operated valves that needed to be manually operated for post fire safe shutdown. The licensee did not recognize that the circuits could cause the valves to become stuck. The team determined licensee personnel would not be able to reposition motor operated valves as specified in plant procedures. The team determined this violation met the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" conditions for receiving enforcement discretion (EA 09 171).
The failure to ensure that safe shutdown equipment could be operated as required during control room fire events was a performance deficiency. The team determined that this finding was more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this violation meets the discretion criteria of the "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)" for a noncompliance identified during the transition to National Fire Protection Association 805, the team determined that discretion to take no enforcement action is appropriate at this time, as described in the Enforcement Policy.
 
Inspection Report# : 2009006 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: FIN Finding Failure to Follow Commitment Tracking Procedures The inspectors identified a finding because the licensee inadvertently deleted procedural steps to recover an emergency diesel generator during a severe accident. The steps were part of a formal commitment to the NRC. The licensee had failed to follow the site commitment management program when making the procedure change and the procedure writer failed to understand the basis for the steps prior to deleting them. The licensee entered this finding in their corrective action program as Condition Reports CR WF3-2009-0193 and CR WF3-2009-1616.
The finding was more than minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, during a severe accident, operators would not have an appropriate mitigation strategy for starting an emergency diesel generator under certain severe accident conditions. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because the finding: (1) could result in a loss of functionality of an emergency diesel generator; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of Human Performance, Decision Making component [H.1(a)], because the licensee failed to use a systematic process when removing the procedural steps Inspection Report# : 2009002 (pdf)
Significance:        Apr 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Voltage Readings Following a Single Cell Battery Charge The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V (Instructions, Procedures and Drawings) because the licensee failed to implement instructions that were intended to help troubleshoot a defective 125 Vdc battery cell. In response to the degraded cell, the licensee had established additional measures to monitor the cell following charging to ensure proper cell operation. However, the licensee did not perform the monitoring. Once identified by the inspectors, the licensee performed more frequent cell tests. The licensee subsequently replaced the faulty cell. The licensee entered this finding into their corrective action program as Condition Reports CR-WF3-2009-1088 and CR-WF3-2009-1099.
The finding was more than minor because it could have resulted in a more significant safety concern if left uncorrected. Specifically, the normal monitoring period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality [P.1(d)]
Inspection Report# : 2009002 (pdf)
Barrier Integrity
 
Emergency Preparedness Significance:      Jul 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Licensee Practices Result in Protective Actions Recommendations for Areas Where Protective Action Guides Are Not Exceeded Green. The inspectors identified a noncited violation of 10 CFR 50.47(b)(10), for the licensees failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Specifically, the licensees guidelines for extending existing protective action recommendations into additional geographical areas of the emergency planning zone under conditions of changing wind vectors were not consistent with the guidance of EPA 400 R 92 001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents." The licensees practices resulted in unnecessary recommendations for protective actions in areas where valid dose projections show federal protective action guides are not exceeded, and may expose members of the public to unjustified risks. The licensee has entered this issue into their corrective action system as Condition Report CR-WF3-2009-03256.
This finding was more than minor because it was not similar to the examples of Manual Chapter 0612, Appendix E, and affected the emergency preparedness cornerstone objective because unnecessary protective actions may expose members of the public to an unjustified risk. The finding was associated with the emergency response organization attributes of 50.47(b) planning standards and training. This finding was of very low safety significance because it was not a risk significant planning standard functional failure or degraded function because licensee protective action recommendations would be issued in accordance with federal guidance for all areas of the emergency planning zone where Protective Action Guides are exceeded. This finding was evaluated as not having a crosscutting aspect because the finding was not indicative of current licensee performance (Section 1EP1).
Inspection Report# : 2009003 (pdf)
Occupational Radiation Safety Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation protection procedural requirements Green. The inspectors reviewed a self-revealing, noncited violation of Technical Specification 6.8.1 which resulted from a worker failing to follow radiation protection procedures. A contract radiation worker went to work near steam generator 1 rather than the area for which he/she was briefed and received multiple electronic dosimeter dose rate alarms, but did not leave the area until receiving a continuous dose alarm. In response, the licensee investigated the occurrence and restricted the individuals access. Additional actions were being evaluated. This issue was entered into the licensees corrective action program as Condition Reports WF3-2009-05648 and WF3-2009-06852.
This finding is greater than minor because it involved the program attribute of exposure control and affected the cornerstone objective in that the failure of the worker to follow procedural guidance resulted in the worker being unknowledgeable to the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not:
(1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a crosscutting aspect in the area of human performance, work practices component, because the worker failed to use human error prevention techniques such as self and peer checking [H.4(a)].
 
Inspection Report# : 2009005 (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 Last modified : May 26, 2010
 
Waterford 3 2Q/2010 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Transient Combustibles (Section 1R05)
The inspectors identified five examples of a green noncited violation of Waterford Steam Electric Station, Unit 3s license condition 2.C.9 for the failure to perform a transient combustible evaluation prior to introducing transient combustibles into procedurally controlled vital plant areas. Specifically, procedures limit the amount of transient combustibles that may be introduced into the control room ventilation equipment room, the component cooling water Train B heat exchanger room, and the main steam isolation valve Train B roof area. Any amounts greater than the preset procedural limits require a transient combustible evaluation to be performed. In all five cases, this evaluation was not performed prior to introduction of the transient combustibles. This violation has been entered into the licensees corrective action program as condition reports CR-WF3-2010-0482, CR-WF3-2010-0598, and CR-WF3-2009-4035.
The performance deficiencies associated with this violation were the failure to comply with Waterford Steam Electric Station, Unit 3s license condition 2.C.9. Specifically, the procedural requirements to perform a transient combustible evaluation prior to introducing the transient combustibles into designated fire zones were not performed. Since several of the previously described fire zones fail to meet 10 CFR50, Appendix R train separation requirements, use of Inspection Manual Chapter 0612, Appendix E to screen for minor examples is not appropriate. This condition is greater than minor because, if left uncorrected, it would become a more significant safety concern, since continued introduction of unevaluated transient combustible loading into controlled areas could significantly reduce the ability to achieve a safe shutdown condition, in the event of a fire in that controlled area. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, to assess the safety significance. Since the severity of the observed deficiencies was assigned a low degradation rating, it was determined to be of very low risk significance. This finding had a crosscutting aspect in the area of human performance associated with the work practices component in that the licensee failed to utilize appropriate human error prevention techniques by (1) discussing transient combustible controls and expectations during pre-job briefs and (2) effectively utilizing human performance barriers, such as posted door signs [H.4(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Update Drawings after Design Change A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for the licensees failure to prescribe an activity affecting quality by documented instructions, procedures, or drawings appropriate to the circumstance. Specifically, for all reactor coolant pump heat exchanger to pump cover bolted connection gasket replacements between the refueling outage of 1986 (RF-1) and the refueling outage of 2009 (RF-16), the licensee prescribed the wrong gasket material, gasket size, and fastener preload because they had failed to incorporate a design change implemented during RF-1 into their instructions, procedures, or drawings. Station modification package SMP-1427, an engineering change implemented during RF-1 in response to industry operating experience, called for a thicker gasket, different gasket material, and an increased bolt preload in order to increase gasket compression and reduce the probability of leakage. As a consequence of failing to incorporate SMP-1427 changes into procedures, all heat exchanger gasket replacements since RF-1, four gasket replacements in total, have utilized thinner gaskets with less than the vendor recommended compression. The licensee documented this condition in CR-WF3-2009-5501.
 
The licensees failure to prescribe appropriate gasket replacement requirements is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available.
This finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee did not institutionalize operating experience through changes to the station procedures
[P.2(b)].
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Reactor Coolant Pump Vapor Seal Leakage A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee did not promptly correct reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on the component cooling water heat exchanger and cover areas of three reactor coolant pumps. Corrective actions for this condition were implemented during refuel outage 15, but these corrective actions failed to correct the condition and the vapor seal leakage continued through operating cycle 16. This resulted in some additional boric acid corrosion and degradation to reactor coolant pump covers and carbon steel component cooling water flanges. The licensee implemented a design modification to correct the condition and documented the condition in CR-WF3-2009-5501.
The licensees failure to promptly correct a condition adverse to quality is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment was still available. This finding had a crosscutting aspect in area of human performance associated with work control in that the licensee did not effectively plan for the resources necessary to implement the post maintenance testing associated with the corrective actions implemented during refuel outage 15, and therefore failed to discover that those corrective actions were inadequate to correct the condition[H.3(a)].
Inspection Report# : 2009005 (pdf)
Significance:      Sep 18, 2009 Identified By: NRC Item Type: FIN Finding Failure to Incorporate Start-Up Transformer Protective Relay Design Basis into Instructions, Procedures, or Drawings.
The team identified a finding for failure to translate design basis criteria into a design basis document for the start-up transformer 3A 51G relay to support the settings listed in Calculation EC E90 012, Protective Relays Settings for Main Generator and Transformers, Revision 1. Without the design basis criteria for the 51G relay, the setpoint values could not be established. Specifically, the team determined that the relay settings listed in Calculation EC E90 012 had not been effectively implemented since the required current transformer ratio of 600/5, upon which the settings were based, was never installed. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04813.
This finding was more than minor because the failure to provide adequate relay setting coordination could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient (initiating event).
The team noted that this finding also applies to 51G relay in the B train which could challenge the single failure criterion. The team determined this finding was of very low safety significance (Green) because the issue would not prevent the safety buses from being reenergized by the emergency diesel generators. Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.10).
 
Inspection Report# : 2009009 (pdf)
Mitigating Systems Significance:        Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify an adverse trend in failures of time-delay relays The team identified a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform a root cause analysis and implement corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, multiple failures of Agastat E7024PB relays that were installed in or designated for safety-related applications constituted a significant condition adverse to quality. The evaluations for the individual relay failures were narrow and did not identify the adverse trend until eight relays had failed in service and seven had failed pre-installation bench tests over a two-year period. The failure of these relays would prevent auto-starting of critical equipment during a loss of offsite power, potentially creating a substantial safety hazard.
The failure of the licensee to recognize that the adverse trend in failures of Agastat E7024PB relays constituted a significant condition adverse to quality, to perform a root cause evaluation, and to initiate corrective actions to prevent recurrence is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the performance deficiency was determined to require a Phase 2 analysis because of the potential for a loss of safety system function. A Phase 2/Phase 3 Significance Determination was performed by an NRC Senior Reactor Analyst. Based on a bounding analysis, the analyst quantitatively determined that the actual change in core damage frequency (?CDF) due to the increased failure rate of Agastat E7024PB relays would be less than 4.0E-7/year. Therefore, this performance deficiency was determined to be of very low safety significance (Green).
This performance deficiency was determined to have a Problem Identification and Resolution cross-cutting aspect in the Corrective Action Program component because the licensee failed to periodically trend and assess information from the Corrective Action Program and other assessments in the aggregate to identify programmatic and common cause problems.
Inspection Report# : 2009010 (pdf)
Significance:        Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate extension of qualified service life of Agastat relays The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which occurred when the licensee inappropriately extended the service life of 322 safety-related Tyco/Agastat series E7000 time-delay relays without having an adequate technical basis. Specifically, the licensees engineering justification for extending the qualified life beyond the manufacturer-recommended ten years considered only degradation due to thermal aging; it failed to consider other known modes of degradation in accordance with applicable industry standards. Further, the team identified that a performance monitoring program intended to assess any increased failure rate due to this change was inappropriately canceled.
The failure of the licensee to perform a complete analysis of aging effects as required by industry standards in extending the qualified life of safety-related Agastat E7000-series relays is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of design control because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this performance deficiency was determined to be of very low safety significance
 
(Green) because it is a design or qualification deficiency confirmed not to result in loss of operability or functionality.
Specifically, only one of the identified relay failures had occurred beyond the recommended 10-year service life; this failure did not result in the failure of multiple redundant trains of safety-related equipment. This finding was determined not to have a cross-cutting aspect because it is not indicative of current licensee performance.
Inspection Report# : 2009010 (pdf)
Significance:        Oct 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow technical specification requirements for Reactor Protective Instrumentation.
The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. The licensee entered this finding in their corrective action program as condition reports CR WF3-2009-4401 and CR-WF3-2009-4407.
The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for reactor protective trips to be inadvertently removed while at power. The failure to meet the technical specifications was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to comply with technical specification 3.3.1 by declaring the log power Channel B inoperable and not placing DNBR, LPD, and reactor coolant flow channels in either bypass or trip condition (H.1.b). (Section 1R15)
Inspection Report# : 2009004 (pdf)
Significance:        Sep 24, 2009 Identified By: Licensee Item Type: VIO Violation Inoperable 125Vdc battery because electricians failed to follow work instructions Following a September 2, 2008 train B 125 Vdc battery failure, the licensee identified a violation of Technical Specification 6.8.1.a for the failure to follow plant procedures during corrective maintenance on the safety-related battery. Following the replacement of the entire battery bank during a 2008 refueling outage, craftsmen identified a faulty battery cell. When replacing the faulty cell, plant workers did not follow all of the specified procedural steps in the work package. The additional work resulted in a loose battery connection that rendered the entire battery bank inoperable. The licensee also failed to address an indicator of the loose connection during the battery discharge test.
The condition then went undetected for several months. The licensee entered this finding in their corrective action program as Condition Report CR WF3 2008-4179.
This finding was greater than minor because it was similar to non-minor example 4.a in NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that the failure to follow site procedures adversely affected safety related equipment. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding required a Phase 2 significance determination because it resulted in the loss of a single train of safety related equipment for greater than the technical specification allowed outage time.
Using a T/2 exposure time of 50 days, the inspectors used the Risk-Informed Inspection Notebook for Waterford Nuclear Power Plant Unit 3, Revision 2.01 and its associated Phase 2 Pre-Solved Table, and determined that a Phase 3 significance determination was necessary. A Region IV senior reactor analyst performed a preliminary Phase 3 significance determination and found that the finding was White. This preliminary Phase 3 significance
 
determination is included as Attachment 2 to this report. This finding had a cross cutting aspect in area of Human Performance (work practices component) because maintenance personnel failed to use appropriate human error prevention techniques, such as peer checking (quality control hold points) and tracking battery components that were loosened (H.4.a). (Section 1R15).
Update: A Regulatory Conference was held for this issue on December 14, 2009. The final significance of this issue was determined to be White as described in a letter to the licensee (ML1001506600), dated January 14, 2010.
Inspection Report# : 2009008 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Account for Reduction of Flow from the Emergency Feedwater System to the Steam Generators The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee did not account for reduction of flow from the emergency feedwater system when analyzing the flow rate to the steam generators and establishing the acceptance criteria for the performance of the motor-driven emergency feedwater pumps. The factors associated with the loss of flow included the emergency diesel generator under-frequency of 0.3 Hertz allowed by technical specifications, and not accounting for accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve. The pumps had a documented analyzed margin of 55 gpm. The margin was reduced by 24 gpm due to allowed diesel under-frequency. Another reduction was attributed to the accepted reverse flow (back leakage) of 25 gpm through the turbine-driven discharge check valve.
This left the combined margin of both emergency feedwater motor-driven pumps at 6 gpm. The licensee entered this issue into the corrective action program as Condition Reports CR-WF3-2009-04731, CR-WF3-2009-04528, and CR-WF3-2009-05043, and performed an operability assessment for each of these factors.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. This finding is of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation. Some margin in total flow still remained to compensate for the reduced pump performance if operated at the reduced-frequency.
The inspectors determined that the finding has a cross cutting aspect in the area of Problem Identification and Resolution, associated with Operating Experience. The licensee had received NRC Information Notice 2008-02, which specifically identified the diesel under-frequency as a potential problem for ac motor-operated pumps, and test acceptance criteria concerns which would have ensured the capability of the equipment to perform its function under the most limiting conditions. The licensee failed to identify the applicability of these potential problems to the emergency feedwater motor-operated pumps and take proper actions [P.2(a)] (Section 1R21.b.1.1).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Proper Design Control Measures to Assure Adequate Design and to Properly Translate the Design into Test Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with three examples.
Example 1: The licensee did not use the correct size emergency feedwater system suction piping in calculation MNQ10-12 Net Positive Suction Head Available for Emergency Feedwater Pumps. The motor-driven pump suction piping is 4 inches in diameter but the licensee nonconservatively used 6-inch piping in the calculations. The licensee has entered this issue into their corrective action program as Condition Report CR-WF3-2009-04729 and performed
 
an operability assessment for the issue.
Example 2: Calculation ECM91-001, Revision 3, Emergency Diesel Generator Fuel Oil Transfer Pump Recirculation and Discharge Flow, arbitrarily assumed that the suction strainer of the fuel oil transfer pump would only be 10 percent clogged. The licensee could not justify the 10 percent clogging assumption or find any justification for selecting the 10 percent value. Also, there is no discussion or any physical comparison to ensure that the mesh of the installed Leslie strainer was the same as that of the Hayward strainer identified in an attachment to the calculation. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04812 and performed an operability assessment for the issue.
Example 3: Calculation EC-I01-003, Revision 0, IST Instrumentation Uncertainties, determines the adequacy of permanent plant instrumentation for inservice testing use. The calculation determined that some specific instruments shall not be used for inservice testing applications. Contrary to the calculation requirements, procedure OP 903 014, used for the inservice testing comprehensive test of the emergency feedwater pumps, specified that the forbidden flow instruments shall be used for verification of emergency feedwater system flow rate. The licensee has entered this issue into their corrective action program as Condition Report CR-W3-2009-04811. These findings are more than minor because they affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance deficiencies would have the potential to lead to more significant safety concerns. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, April 20 23, 2009, that included the emergency feedwater turbine-driven pump and the emergency diesel generator fuel oil transfer pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify any of these three issues [P.3.(a)] (Section 1R21.b.1.6).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to have an Operating Procedure for Executing an Evolution Credited in the UFSAR and in an Request for a License Amendment The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings pertaining to the emergency diesel generator fuel oil transfer pump. Criterion V states, in part, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Specifically, the licensee did not have operating procedures for accomplishing the transfer of fuel oil from one storage tank to the opposite train feed tank (day tank) using the opposite train fuel oil transfer pump, as designated in the USAR Table 9.5-2, Failure Mode and Effects Analysis. Also, License Amendment Number 157 (TAC Number MA4940) was granted, in part, for having the capability to transfer fuel oil from one storage tank to the opposite train feed tank using the opposite transfer pump. The licensee specified this capability as part of the justification for having an insufficiently sized fuel oil storage tank. Moreover, the Safety Evaluation Report associated with License Amendment Number 157 specifically referred to this capability at Waterford 3, and specified that procedures were available for accomplishing the transfer of fuel oil. The licensee has entered this finding in their corrective action program as Condition Report CR-WF3-2009-04950, and performed an operability assessment for the issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events.
Also, using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Section 1-3, Screen for More than Minor - ROP, question 2, the finding is more than minor because if left uncorrected, the performance
 
deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, the finding was determined to have very low safety significance (Green) because the failure to have an operating procedure did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.7).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Analyze the Effect of Acceptable Reverse Flow through Emergency Feedwater Check Valves The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to analyze the effects of the acceptable back leakage of 25 gpm from the emergency feedwater pump discharge check valves on the integrity of the emergency feedwater pumps and the integrity of its suction piping. The acceptable back leakage could possibly cause the pump to reverse rotate, and provide a path for high pressure fluid to go through the pump and pressurize low pressure suction piping. The licensee has entered this item in their corrective action program as Condition Report CR WF3 2009 04528 and performed an operability assessment for this issue.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events. This finding closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. This finding was determined to be of very low safety significance (Green) because this design issue did not result in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the Technical Specification allowed outage time, and did not affect external event mitigation.
The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessment. The licensee conducted a Waterford 3 Component Design Basis Assessment, on April 20-23, 2009, that included the emergency feedwater AB turbine-driven pump in the Scope of Components to be Reviewed During CDBI Assessment, and failed to identify the impact of reverse flow on the integrity of the pump and its suction piping [P.3.(a)] (Section 1R21.b.1.8).
Inspection Report# : 2009009 (pdf)
Significance:        Sep 18, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify or Check the adequacy of Design Changes for the Emergency Diesel Generator Protective Relay IGVC-51V The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. The calculation EE2 14 3 Diesel Generator Overcurrent Protection, Revision 1, does not document sufficient design bases for the setting of the IGCV 51 overcurrent with voltage control relays for the emergency diesel generators.
Specifically, the licensee failed to perform an adequate evaluation of new setpoint values identified in Engineering Report ER W3 99 0174 00 00, which provided the bases for relay tap setpoint changes for emergency diesel generator overcurrent protection while the diesel was in test mode. The primary purpose of the IGCV-51V relays was to protect the emergency diesel generator against external faults and prevent the output breaker from closing following a breaker trip associated with a fault. If the faulted bus had been isolated by the operation of the under-voltage relays instead of the IGCV 51 relays, the emergency diesel generator output breaker would be allowed to electrically reclose onto this faulted bus and potentially damage the emergency diesel generator and the associated switchgear. The issue has been entered into the licensees corrective action program as Condition Report CR WF3 2009 04780.
 
The failure to have sufficient design bases for the emergency diesel generator overcurrent protection IGCV 51V relays without an adequate verification of the setpoint modification for the IGCV 51V relay, Voltage Controlled, Time-Overcurrent Relay, for emergency diesel generator overcurrent protection while the diesel was in test mode, was a performance deficiency. Specifically, failure to verify the adequacy of a design modification for the IGCV 51V relay could result in reduced reliability of the emergency diesel generators. The finding was determined to be greater than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences.
Using Manual Chapter 0609.04, the finding was determined to have a very low safety significance (Green) because the failure did not result in loss of operability or functionality and because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding was reviewed for crosscutting aspects and none were identified (Section 1R21.b.1.12).
Inspection Report# : 2009009 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation protection procedural requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 6.8.1 which resulted from a worker failing to follow radiation protection procedures. A contract radiation worker went to work near steam generator 1 rather than the area for which he/she was briefed and received multiple electronic dosimeter dose rate alarms, but did not leave the area until receiving a continuous dose alarm. In response, the licensee investigated the occurrence and restricted the individuals access. Additional actions were being evaluated. This issue was entered into the licensees corrective action program as Condition Reports CR WF3-2009-05648 and CR WF3-2009-06852.
This finding is greater than minor because it involved the program attribute of exposure control and affected the cornerstone objective in that the failure of the worker to follow procedural guidance resulted in the worker being unknowledgeable to the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not:
(1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a crosscutting aspect in the area of human performance, work practices component, because the worker failed to use human error prevention techniques such as self and peer checking [H.4(a)].
Inspection Report# : 2009005 (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 Last modified : September 02, 2010
 
Waterford 3 3Q/2010 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Transient Combustibles (Section 1R05)
The inspectors identified five examples of a green noncited violation of Waterford Steam Electric Station, Unit 3s license condition 2.C.9 for the failure to perform a transient combustible evaluation prior to introducing transient combustibles into procedurally controlled vital plant areas. Specifically, procedures limit the amount of transient combustibles that may be introduced into the control room ventilation equipment room, the component cooling water Train B heat exchanger room, and the main steam isolation valve Train B roof area. Any amounts greater than the preset procedural limits require a transient combustible evaluation to be performed. In all five cases, this evaluation was not performed prior to introduction of the transient combustibles. This violation has been entered into the licensees corrective action program as condition reports CR-WF3-2010-0482, CR-WF3-2010-0598, and CR-WF3-2009-4035.
The performance deficiencies associated with this violation were the failure to comply with Waterford Steam Electric Station, Unit 3s license condition 2.C.9. Specifically, the procedural requirements to perform a transient combustible evaluation prior to introducing the transient combustibles into designated fire zones were not performed. Since several of the previously described fire zones fail to meet 10 CFR50, Appendix R train separation requirements, use of Inspection Manual Chapter 0612, Appendix E to screen for minor examples is not appropriate. This condition is greater than minor because, if left uncorrected, it would become a more significant safety concern, since continued introduction of unevaluated transient combustible loading into controlled areas could significantly reduce the ability to achieve a safe shutdown condition, in the event of a fire in that controlled area. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, to assess the safety significance. Since the severity of the observed deficiencies was assigned a low degradation rating, it was determined to be of very low risk significance. This finding had a crosscutting aspect in the area of human performance associated with the work practices component in that the licensee failed to utilize appropriate human error prevention techniques by (1) discussing transient combustible controls and expectations during pre-job briefs and (2) effectively utilizing human performance barriers, such as posted door signs [H.4(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Update Drawings after Design Change A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for the licensees failure to prescribe an activity affecting quality by documented instructions, procedures, or drawings appropriate to the circumstance. Specifically, for all reactor coolant pump heat exchanger to pump cover bolted connection gasket replacements between the refueling outage of 1986 (RF-1) and the refueling outage of 2009 (RF-16), the licensee prescribed the wrong gasket material, gasket size, and fastener preload because they had failed to incorporate a design change implemented during RF-1 into their instructions, procedures, or drawings. Station modification package SMP-1427, an engineering change implemented during RF-1 in response to industry operating experience, called for a thicker gasket, different gasket material, and an increased bolt preload in order to increase gasket compression and reduce the probability of leakage. As a consequence of failing to incorporate SMP-1427 changes into procedures, all heat exchanger gasket replacements since RF-1, four gasket replacements in total, have utilized thinner gaskets with less than the vendor recommended compression. The licensee documented this condition in CR-WF3-2009-5501.
 
The licensees failure to prescribe appropriate gasket replacement requirements is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available.
This finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee did not institutionalize operating experience through changes to the station procedures
[P.2(b)].
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Reactor Coolant Pump Vapor Seal Leakage A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee did not promptly correct reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on the component cooling water heat exchanger and cover areas of three reactor coolant pumps. Corrective actions for this condition were implemented during refuel outage 15, but these corrective actions failed to correct the condition and the vapor seal leakage continued through operating cycle 16. This resulted in some additional boric acid corrosion and degradation to reactor coolant pump covers and carbon steel component cooling water flanges. The licensee implemented a design modification to correct the condition and documented the condition in CR-WF3-2009-5501.
The licensees failure to promptly correct a condition adverse to quality is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment was still available. This finding had a crosscutting aspect in area of human performance associated with work control in that the licensee did not effectively plan for the resources necessary to implement the post maintenance testing associated with the corrective actions implemented during refuel outage 15, and therefore failed to discover that those corrective actions were inadequate to correct the condition[H.3(a)].
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:      May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Repetitive Voiding in the Low Pressure Safety Injection System The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to preclude repetition of a significant condition adverse to quality. Specifically, licensee corrective actions to prevent recurrence of voiding in the low pressure safety injection system were not sufficient to prevent nitrogen voids from challenging system operability. This violation was entered into the licensees corrective action program as CR WF3 2010 3050.
The finding is more than minor because, if left uncorrected, the finding would have the potential to become a more significant safety concern (i.e., continued challenges to system operability). Using Manual Chapter 0609.04, Phase 1
- Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in
 
the corrective action component of the problem identification and resolution area in that the licensee failed to thoroughly evaluate the problem, such that the resolutions addressed the cause [P.1(c)]. As a result, the resolutions failed to prevent recurrence of the problem (Section 4OA2.5a).
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Technical Specification 3.7.5 Action Statement The team identified a non-cited violation of 10 CFR 50.36 (b), Technical Specifications, for failure to derive technical specifications from the analyses and evaluation included in the safety analysis report. Specifically, the licensee failed to derive an action statement for Technical Specification 3.7.5 that meets the assumptions included in the Waterford Unit 3 Updated Safety Analysis Report. The Updated Safety Analysis Report evaluation assumes an instantaneous levee failure occurs at a Mississippi River level of +27 feet mean sea level. The inspectors determined that the action statement for Technical Specification 3.7.5, to complete procedures to secure doors and penetrations in 12 hours, was not derived from the evaluation included in the safety analysis report because the actions would take place after the assumed instantaneous levee failure. The licensee entered this condition into the corrective action program as CR WF3 2010 03232. As a short term compensatory measure, the licensee established criteria for taking appropriate action before the Mississippi River level would reach the +27 feet mean sea level safety limit.
The finding is more than minor because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. In addition, the performance deficiency adversely affects the Mitigating Systems Cornerstone attribute of external events to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the finding was of very low safety significance (Green) because it was a nonconforming condition that did not result in complete unavailability of the equipment (Section 4OA2.5b).
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to ensure that conditions adverse to quality are promptly corrected. Specifically, multiple examples of boric acid leaks were identified in the corrective action program where corrective actions had not yet been taken or had been ineffective. At least ten of these active boric acid leaks are five to seven years old.
The failure to promptly correct boric acid leaks is a performance deficiency. The finding is more than minor because, if left uncorrected, the finding could become a more significant safety concern (i.e., potential for damage to carbon steel components or inhibiting the safety-function of others). Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the problem identification and resolution, corrective action component [P.1(d)] in that the licensee failed to effectively correct identified boric acid leaks in a timely manner (Section 4OA2.5c).
Inspection Report# : 2010006 (pdf)
Significance:        Mar 04, 2010 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Faulty Relays A self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because Entergy did not conduct timely corrective actions to preclude repetition of a significant condition adverse to quality that involved Tyco relay replacements. Specifically, Entergy extended the due date of corrective actions to preclude repetition of suspected faulty relays without an adequate justification. As a result, this led to additional relay failures that challenged the reliability of risk significance safety systems. Entergy entered this issue into their corrective action program for resolution as condition reports CR-WF3-2010-1330 and CR-WF3-2010-4199. The immediate corrective actions after the additional failures included initiating work requests to replace the faulty relays. The planned corrective actions included an evaluation of the effectiveness and timeliness of the Tyco replacement project.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, Entergy did not provide an adequate justification to extend corrective actions beyond its original due date such that it could not affect the availability, reliability, and capability of risk significance safety systems. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it Technical Specification allowed outage time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the corrective action component of problem identification and resolution area because Entergy did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity [P.1.d] (Section 4OA2).
Inspection Report# : 2010003 (pdf)
Significance:        Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify an adverse trend in failures of time-delay relays The team identified a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform a root cause analysis and implement corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, multiple failures of Agastat E7024PB relays that were installed in or designated for safety-related applications constituted a significant condition adverse to quality. The evaluations for the individual relay failures were narrow and did not identify the adverse trend until eight relays had failed in service and seven had failed pre-installation bench tests over a two-year period. The failure of these relays would prevent auto-starting of critical equipment during a loss of offsite power, potentially creating a substantial safety hazard.
The failure of the licensee to recognize that the adverse trend in failures of Agastat E7024PB relays constituted a significant condition adverse to quality, to perform a root cause evaluation, and to initiate corrective actions to prevent recurrence is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the performance deficiency was determined to require a Phase 2 analysis because of the potential for a loss of safety system function. A Phase 2/Phase 3 Significance Determination was performed by an NRC Senior Reactor Analyst. Based on a bounding analysis, the analyst quantitatively determined that the actual change in core damage frequency (?CDF) due to the increased failure rate of Agastat E7024PB relays would be less than 4.0E-7/year. Therefore, this performance deficiency was determined to be of very low safety significance (Green).
This performance deficiency was determined to have a Problem Identification and Resolution cross-cutting aspect in the Corrective Action Program component because the licensee failed to periodically trend and assess information from the Corrective Action Program and other assessments in the aggregate to identify programmatic and common cause problems.
Inspection Report# : 2009010 (pdf)
 
Significance:      Oct 19, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate extension of qualified service life of Agastat relays The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which occurred when the licensee inappropriately extended the service life of 322 safety-related Tyco/Agastat series E7000 time-delay relays without having an adequate technical basis. Specifically, the licensees engineering justification for extending the qualified life beyond the manufacturer-recommended ten years considered only degradation due to thermal aging; it failed to consider other known modes of degradation in accordance with applicable industry standards. Further, the team identified that a performance monitoring program intended to assess any increased failure rate due to this change was inappropriately canceled.
The failure of the licensee to perform a complete analysis of aging effects as required by industry standards in extending the qualified life of safety-related Agastat E7000-series relays is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of design control because it affects the availability and reliability of systems which respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this performance deficiency was determined to be of very low safety significance (Green) because it is a design or qualification deficiency confirmed not to result in loss of operability or functionality.
Specifically, only one of the identified relay failures had occurred beyond the recommended 10-year service life; this failure did not result in the failure of multiple redundant trains of safety-related equipment. This finding was determined not to have a cross-cutting aspect because it is not indicative of current licensee performance.
Inspection Report# : 2009010 (pdf)
Significance:      Oct 07, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow technical specification requirements for Reactor Protective Instrumentation.
The inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. The licensee entered this finding in their corrective action program as condition reports CR WF3-2009-4401 and CR-WF3-2009-4407.
The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for reactor protective trips to be inadvertently removed while at power. The failure to meet the technical specifications was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to comply with technical specification 3.3.1 by declaring the log power Channel B inoperable and not placing DNBR, LPD, and reactor coolant flow channels in either bypass or trip condition (H.1.b). (Section 1R15)
Inspection Report# : 2009004 (pdf)
Barrier Integrity
 
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation protection procedural requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 6.8.1 which resulted from a worker failing to follow radiation protection procedures. A contract radiation worker went to work near steam generator 1 rather than the area for which he/she was briefed and received multiple electronic dosimeter dose rate alarms, but did not leave the area until receiving a continuous dose alarm. In response, the licensee investigated the occurrence and restricted the individuals access. Additional actions were being evaluated. This issue was entered into the licensees corrective action program as Condition Reports CR WF3-2009-05648 and CR WF3-2009-06852.
This finding is greater than minor because it involved the program attribute of exposure control and affected the cornerstone objective in that the failure of the worker to follow procedural guidance resulted in the worker being unknowledgeable to the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not:
(1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a crosscutting aspect in the area of human performance, work practices component, because the worker failed to use human error prevention techniques such as self and peer checking [H.4(a)].
Inspection Report# : 2009005 (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 Last modified : November 29, 2010
 
Waterford 3 4Q/2010 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Transient Combustibles The inspectors identified five examples of a green noncited violation of Waterford Steam Electric Station, Unit 3s license condition 2.C.9 for the failure to perform a transient combustible evaluation prior to introducing transient combustibles into procedurally controlled vital plant areas. Specifically, procedures limit the amount of transient combustibles that may be introduced into the control room ventilation equipment room, the component cooling water Train B heat exchanger room, and the main steam isolation valve Train B roof area. Any amounts greater than the preset procedural limits require a transient combustible evaluation to be performed. In all five cases, this evaluation was not performed prior to introduction of the transient combustibles. This violation has been entered into the licensees corrective action program as condition reports CR-WF3-2010-0482, CR-WF3-2010-0598, and CR-WF3-2009-4035.
The performance deficiencies associated with this violation were the failure to comply with Waterford Steam Electric Station, Unit 3s license condition 2.C.9. Specifically, the procedural requirements to perform a transient combustible evaluation prior to introducing the transient combustibles into designated fire zones were not performed. Since several of the previously described fire zones fail to meet 10 CFR50, Appendix R train separation requirements, use of Inspection Manual Chapter 0612, Appendix E to screen for minor examples is not appropriate. This condition is greater than minor because, if left uncorrected, it would become a more significant safety concern, since continued introduction of unevaluated transient combustible loading into controlled areas could significantly reduce the ability to achieve a safe shutdown condition, in the event of a fire in that controlled area. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, to assess the safety significance. Since the severity of the observed deficiencies was assigned a low degradation rating, it was determined to be of very low risk significance. This finding had a crosscutting aspect in the area of human performance associated with the work practices component in that the licensee failed to utilize appropriate human error prevention techniques by (1) discussing transient combustible controls and expectations during pre-job briefs and (2) effectively utilizing human performance barriers, such as posted door signs.
Inspection Report# : 2010002 (pdf)
Mitigating Systems Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not promptly correct a condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of degraded diodes in static uninterruptible power supplies A and B, respectively. As a result, this led to another failure of the static uninterruptible power supply A. The licensee entered this issue into their corrective action program (CAP) for resolution as CR-WF3-2010-6760. The immediate corrective actions following the additional failure included installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier. The planned corrective actions included implementation of an
 
increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers.
This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations given the importance of the system and the fact that this condition affects both static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it Technical Specification allowed outage time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the corrective action component of problem identification and resolution area because the licensee did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2010005 (pdf)
Significance:        Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Faulty Relays A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not conduct timely corrective actions to preclude repetition of a significant condition adverse to quality that involved Tyco relay replacements. Specifically, the licensee extended the due date of corrective actions to preclude repetition of suspected faulty relays without an adequate justification. As a result, this led to additional relay failures that challenged the reliability of risk significance safety systems. The immediate corrective actions after the additional failures included initiating work requests to replace the faulty relays. The planned corrective actions included an evaluation of the effectiveness and timeliness of the Tyco replacement project. The licensee entered this issue into their corrective action program for resolution as Condition Reports CR WF3 2010 1330 and CR WF3 2010 4199.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not provide an adequate justification to extend corrective actions beyond its original due date such that it could not affect the availability, reliability, and capability of risk significance safety systems. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding was determined to have very low safety significance because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to external events. The finding has a crosscutting aspect in the corrective action component of problem identification and resolution area because the licensee did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2010003 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Repetitive Voiding in the Low Pressure Safety Injection System The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to preclude repetition of a significant condition adverse to quality. Specifically, licensee corrective actions to prevent recurrence of voiding in the low pressure safety injection system were not sufficient to prevent nitrogen voids from challenging system operability. This violation was entered into the licensees corrective action program as CR WF3 2010 3050.
 
The finding is more than minor because, if left uncorrected, the finding would have the potential to become a more significant safety concern (i.e., continued challenges to system operability). Using Manual Chapter 0609.04, Phase 1
- Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the corrective action component of the problem identification and resolution area in that the licensee failed to thoroughly evaluate the problem, such that the resolutions addressed the cause. As a result, the resolutions failed to prevent recurrence of the problem.
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Technical Specification 3.7.5 Action Statement The team identified a noncited violation of 10 CFR 50.36 (b), Technical Specifications, for failure to derive technical specifications from the analyses and evaluation included in the safety analysis report. Specifically, the licensee failed to derive an action statement for Technical Specification 3.7.5 that meets the assumptions included in the Waterford Unit 3 Updated Safety Analysis Report. The Updated Safety Analysis Report evaluation assumes an instantaneous levee failure occurs at a Mississippi River level of +27 feet mean sea level. The inspectors determined that the action statement for Technical Specification 3.7.5, to complete procedures to secure doors and penetrations in 12 hours, was not derived from the evaluation included in the safety analysis report because the actions would take place after the assumed instantaneous levee failure. The licensee entered this condition into the corrective action program as CR WF3 2010 03232. As a short term compensatory measure, the licensee established criteria for taking appropriate action before the Mississippi River level would reach the +27 feet mean sea level safety limit.
The finding is more than minor because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. In addition, the performance deficiency adversely affects the Mitigating Systems Cornerstone attribute of external events to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the finding was of very low safety significance (Green) because it was a nonconforming condition that did not result in complete unavailability of the equipment.
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Quality The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to ensure that conditions adverse to quality are promptly corrected. Specifically, multiple examples of boric acid leaks were identified in the corrective action program where corrective actions had not yet been taken or had been ineffective. At least ten of these active boric acid leaks are five to seven years old.
The failure to promptly correct boric acid leaks is a performance deficiency. The finding is more than minor because, if left uncorrected, the finding could become a more significant safety concern (i.e., potential for damage to carbon steel components or inhibiting the safety-function of others). Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the
 
problem identification and resolution, corrective action component in that the licensee failed to effectively correct identified boric acid leaks in a timely manner.
Inspection Report# : 2010006 (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 Last modified : March 03, 2011
 
Waterford 3 1Q/2011 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Initiating Events Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not adequately implement the operability determination process requirements in accordance with EN-OP-104, Operability Determination Process. Specifically, the licensee did not monitor a degraded and non-conformance condition associated with the reactor coolant pump N-9000 stage seals as required by EN-OP-104. As a result, the licensee did not perform a new operability determination after assumptions and compensatory measures identified in the original operability determination changed. This also led to compliance issues with technical specifications and missed maintenance rule functional failures. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1965. The immediate corrective actions included revising the operability determination to account for the current configuration. The planned corrective actions included the licensee replacing the degraded reactor coolant pump seals during the next two refueling outages.
The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition that had the potential to lead to a small loss of coolant accident. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification completion time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not promptly correct a condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of degraded diodes in static uninterruptible power supplies A and B, respectively. As a result, this led to another failure of the static uninterruptible power supply A. The licensee entered this issue into their corrective action program (CAP) for resolution as CR-WF3-2010-6760. The immediate corrective actions following the additional failure included installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier. The planned corrective actions included implementation of an
 
increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers.
This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations given the importance of the system and the fact that this condition affects both static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it Technical Specification allowed outage time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the corrective action component of problem identification and resolution area because the licensee did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2010005 (pdf)
Significance:        Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Faulty Relays A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not conduct timely corrective actions to preclude repetition of a significant condition adverse to quality that involved Tyco relay replacements. Specifically, the licensee extended the due date of corrective actions to preclude repetition of suspected faulty relays without an adequate justification. As a result, this led to additional relay failures that challenged the reliability of risk significance safety systems. The immediate corrective actions after the additional failures included initiating work requests to replace the faulty relays. The planned corrective actions included an evaluation of the effectiveness and timeliness of the Tyco replacement project. The licensee entered this issue into their corrective action program for resolution as Condition Reports CR WF3 2010 1330 and CR WF3 2010 4199.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not provide an adequate justification to extend corrective actions beyond its original due date such that it could not affect the availability, reliability, and capability of risk significance safety systems. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding was determined to have very low safety significance because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to external events. The finding has a crosscutting aspect in the corrective action component of problem identification and resolution area because the licensee did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2010003 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Repetitive Voiding in the Low Pressure Safety Injection System The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to preclude repetition of a significant condition adverse to quality. Specifically, licensee corrective actions to prevent recurrence of voiding in the low pressure safety injection system were not sufficient to prevent nitrogen voids from challenging system operability. This violation was entered into the licensees corrective action program as CR WF3 2010 3050.
 
The finding is more than minor because, if left uncorrected, the finding would have the potential to become a more significant safety concern (i.e., continued challenges to system operability). Using Manual Chapter 0609.04, Phase 1
- Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the corrective action component of the problem identification and resolution area in that the licensee failed to thoroughly evaluate the problem, such that the resolutions addressed the cause. As a result, the resolutions failed to prevent recurrence of the problem.
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Derive Technical Specifications from Analysis The team identified a noncited violation of 10 CFR 50.36 (b), Technical Specifications, for failure to derive technical specifications from the analyses and evaluation included in the safety analysis report. Specifically, the licensee failed to derive an action statement for Technical Specification 3.7.5 that meets the assumptions included in the Waterford Unit 3 Updated Safety Analysis Report. The Updated Safety Analysis Report evaluation assumes an instantaneous levee failure occurs at a Mississippi River level of +27 feet mean sea level. The inspectors determined that the action statement for Technical Specification 3.7.5, to complete procedures to secure doors and penetrations in 12 hours, was not derived from the evaluation included in the safety analysis report because the actions would take place after the assumed instantaneous levee failure. The licensee entered this condition into the corrective action program as CR WF3 2010 03232. As a short term compensatory measure, the licensee established criteria for taking appropriate action before the Mississippi River level would reach the +27 feet mean sea level safety limit.
The finding is more than minor because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. In addition, the performance deficiency adversely affects the Mitigating Systems Cornerstone attribute of external events to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the finding was of very low safety significance (Green) because it was a nonconforming condition that did not result in complete unavailability of the equipment.
Inspection Report# : 2010006 (pdf)
Significance:        May 28, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Multiple Conditions Adverse to Quality The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to ensure that conditions adverse to quality are promptly corrected. Specifically, multiple examples of boric acid leaks were identified in the corrective action program where corrective actions had not yet been taken or had been ineffective. At least ten of these active boric acid leaks are five to seven years old.
The failure to promptly correct boric acid leaks is a performance deficiency. The finding is more than minor because, if left uncorrected, the finding could become a more significant safety concern (i.e., potential for damage to carbon steel components or inhibiting the safety-function of others). Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, the issue screened as having very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not represent a loss of risk significant non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the
 
problem identification and resolution, corrective action component in that the licensee failed to effectively correct identified boric acid leaks in a timely manner.
Inspection Report# : 2010006 (pdf)
Barrier Integrity Significance:      Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Barrier Integrity Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not conduct required technical specification surveillance testing on equipment in an as-found condition. Specifically, the licensee performed corrective maintenance (preconditioning) on the system to achieve more favorable results, prior to completing the surveillance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1927. The immediate correction action included the performance of the control room envelop tracer gas test.
The finding is more than minor because it is associated with the barrier performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not properly perform testing on equipment to evaluate barrier performance. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating the need for planned contingencies, compensatory actions, and abort criteria.
Inspection Report# : 2011002 (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 07, 2011
 
Waterford 3 2Q/2011 Plant Inspection Findings Initiating Events Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the FSAR following Modifications to the Reactor Coolant Pump Vapor Seals.
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) because the licensee did not revise the final safety analysis report (FSAR) as updated with information consistent with plant conditions.
Specifically, the licensee did not update Section 5.4.1.3 of the FSAR for Waterford Steam Electric Station, Unit 3 following modifications to the reactor coolant pump vapor seals in 2007 and 2009, respectively. As a result, the licensee did not promptly identify and correct FSAR noncompliance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7421. The planned corrective actions include revising the FSAR as updated and replacing the degraded reactor coolant pump seals during the next two refueling outages.
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because the longstanding and incorrect information in the FSAR as updated had a material impact on safety and licensed activities. The material impact is that the modifications created a reactor coolant pump seal loss of coolant accident likelihood inside containment, which could have potentially impacted licensed activities. The inspectors determined the violation is a Severity Level IV (very low safety significance) since the erroneous information not updated in the FSAR was not used to make an unacceptable change to the facility nor impacted a licensing or safety decision by the NRC. The inspectors determined there is a cross-cutting aspect in the corrective action component of the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate and take adequate actions in a timely manner to update the FSAR to be consistent with plant conditions [P.1.c of IMC 0310] (Section 1R18).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Work Order Instructions to Restore a Feedwater Heater Drain Valve.
A self-revealing finding occurred because maintenance personnel did not follow written procedures during the calibration of a level switch that controls feedwater heater drain valve FHD703A. Specifically, the licensee did not perform concurrent verification checks as required by documented work order instructions (WO-00180716) to ensure that personnel restore manipulate components to the correct position following maintenance. As a result, the feedwater heater drain valve remained in a closed manipulate state, which caused a spurious isolation of a string of feedwater heaters. The isolation of the feedwater heaters caused operators to down power the reactor to approximately 72 percent. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2009-7420.
The immediate corrective actions included restoring the feedwater heater drain valve to its proper position.
The finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the human error caused an event that upset plant stability during power operation. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it does not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensees personnel proceed in the face
 
of uncertainty or unexpected circumstances [H.4.a of IMC 0310] (Section 4OA2.3).
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Determination Process for a Degraded and Non-Conforming condition Related to Reactor Coolant Pump N9000 Seals The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not adequately implement the operability determination process requirements in accordance with EN-OP-104, Operability Determination Process. Specifically, the licensee did not monitor a degraded and non-conformance condition associated with the reactor coolant pump N-9000 stage seals as required by EN-OP-104. As a result, the licensee did not perform a new operability determination after assumptions and compensatory measures identified in the original operability determination changed. This also led to compliance issues with technical specifications and missed maintenance rule functional failures. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1965. The immediate corrective actions included revising the operability determination to account for the current configuration. The planned corrective actions included the licensee replacing the degraded reactor coolant pump seals during the next two refueling outages.
The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition that had the potential to lead to a small loss of coolant accident. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification completion time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Monitor Activities Associated with the Internal Conditions of the Condensate and Refueling Water Storage Pool Structures.
The inspectors identified a non-cited violation of 10 CFR 50.65(a)(3) because the licensee did not evaluate or adequately monitor activities associated with the condition of the condensate and refueling water storage pools structures. Specifically, the licensee did not evaluate the internal condition of the storage pools through the performance of appropriate preventive maintenance activities and did not evaluate these activities at least every refueling cycle, where practical, for industry-wide operating experience. As a result, there is no preventive maintenance developed for this activity when previous industry-wide operating experience documented previous issues of concrete deterioration due to contact with boric acid over a long period of time. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1168. The planned corrective actions include the development of appropriate preventive maintenance activities to examine the internal conditions of the storage pool structures during the refuel outages.
 
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, with no preventive maintenance to monitor the internal conditions of the storage pools, this would impact the reliability of the structures. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee did not implement and institutionalizes operating experience through changes to station processes, procedures, equipment, and training programs [P.2.b of IMC 0310] (Section 1R12).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Written Procedures for Restoring a Time Delay Relay Associated with the 'A' Emergency Diesel Generator Output Breaker.
A self-revealing non-cited violation of Technical Specification 6.8.1.a occurred because the licensee did not implement written procedures and instructions. Specifically, maintenance personnel did not follow procedure ME-007-005, Time Delay Relay Setting Check, Adjustment, and Functional Test, during the lifting leads process for restoration of a time delay relay (EG EREL2327-C) associated with the A emergency diesel generator (EDG) maintenance activity. As a result, the A EDG output breaker did not automatically close during technical specification surveillance testing because the leads on the relay were wired incorrectly. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3190. The immediate corrective action included the re-wiring of the relay.
The finding is more than minor because it is associated with the human and equipment performance attributes of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the licensee did not ensure the availability, reliability and capability of the A EDG through human error prevention techniques. The senior resident inspector performed the initial significance determination for the diesel generator output breaker failure. The inspector used the NRC IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. The senior reactor analyst considered the output breaker a part of the emergency diesel generator component boundary. Assuming a one year exposure period, the finding was potentially Yellow, which warranted further review.
Therefore, the senior reactor analyst performed a bounding Phase 3 significance determination. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was approximately 5.4E-7/year. The dominant core damage sequences included loss of offsite power events, failure of the output breaker recovery action, independent failure of the other emergency diesel generator and failure to recover offsite power in 4 hours. Equipment that helped mitigate the risk included the ability of an operator to recover the output breaker. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not communicate human performance error prevention techniques, such as self and peer checking, and proper documentation of activities [H.4.a of IMC 0310] (Section 1R19).
Inspection Report# : 2011003 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not promptly correct a condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of degraded diodes in static uninterruptible power supplies A and B, respectively. As a result, this led to another failure of the static uninterruptible power supply A. The licensee entered this issue into their corrective action program (CAP) for resolution as CR-WF3-2010-6760. The immediate corrective actions following the additional failure included installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier. The planned corrective actions included implementation of an increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers.
This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations given the importance of the system and the fact that this condition affects both static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it Technical Specification allowed outage time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the decision-making component of human performance because the licensee did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained.
Inspection Report# : 2010005 (pdf)
Barrier Integrity Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Control Room Envelope Preconditioning The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not conduct required technical specification surveillance testing on equipment in an as-found condition. Specifically, the licensee performed corrective maintenance (preconditioning) on the system to achieve better results, prior to completing the surveillance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1927. The immediate corrective action included the performance of the control room envelope tracer gas test.
The finding is more than minor because it is associated with the barrier performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not properly perform testing on equipment to evaluate barrier performance. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding doesnt represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room. The finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating the need for planned contingencies, compensatory actions, and abort criteria [H.3.a of IMC 0310] (Section 1R22).
Inspection Report# : 2011002 (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 : October 14, 2011
 
Waterford 3 3Q/2011 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement a Reactor Coolant System Drain Down Procedure The inspectors documented a self-revealing non-cited violation of Technical Specification 6.8.1.a because the licensee did not adequately implement Operating Procedure OP-001-003, Reactor Coolant System Drain Down, during the installation of the incore instrumentation flanges. Specifically, the licensee did not establish a reactor coolant system vent path while maintaining reactor coolant level below 26 feet for the assembly of the incore instrumentation flanges as required by OP-001-003. As a result, the licensee experienced a loss of reactor coolant inventory from three unassembled incore instrumentation flanges, which spilled onto the reactor vessel head insulation and filled the upper annulus cavity of the reactor vessel. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3163 and CR-WF3-2011-3636. The immediate corrective actions included opening the pressurizer spray line vent valve (RC-309) to establish a reactor coolant system vent path.
The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination for the failure to adequately implement operating procedures using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on the conditions of the plant at the time of the event. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately coordinate work activities in incorporating actions to address the impact of the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities
[H.3(b)]. (Section 1R20.1)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Testing for a Shutdown Cooling Heat Exchanger Outlet Stop Check Valve The inspectors documented a self-revealing non-cited violation of 10 CFR 50.55a, Codes and Standards, because the licensee did not establish and maintain an adequate testing program for a shutdown cooling heat exchanger outlet stop check valve (CS-117A) in accordance with Mandatory Appendix II, Check Valve Condition Monitoring Program, of the American Society of Mechanical Engineers Operation and Maintenance Code 2001 through 2003.
Specifically, the licensee did not provide adequate inservice testing to detect degradation of seat leakage on the stop check valve CS-117A. As a result, the operating train of shutdown cooling experienced a flow diversion when the licensee opened the upstream containment spray isolation header valve to fill the containment spray riser. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3350 and CR-WF3-2011-5841.
The immediate corrective action included the closure of the upstream isolation valve and the initiation of a work order to address seat leakage on the stop check valve CS-117. The planned corrective action includes the development of an augmented test to determine appropriate seat leakage criteria for the stop check valve.
 
The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 -
Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, since the degraded stop check valve upsets plant stability and challenge critical safety functions during shutdown conditions. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding did not have a cross-cutting aspect associated with it because the licensee established the check valve condition monitoring program prior to the past three years. Therefore it is not reflective of current plant performance. (Section 1R20.2)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Apparent Cause Evaluation Process Procedure The inspectors identified a finding because the licensee did not implement procedure EN-LI-119, Apparent Cause Evaluation Process. Specifically, the licensee did not follow the requirements provided in procedure EN-LI-119, Section 5.3.3 (k), to complete corrective actions in a timely manner for the intersystem leakage in the gas waste management system. As a result, no corrective action implementation occurred prior to additional equipment failures for the system. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0934. The immediate corrective action included the reevaluation of the causal determination and development of an implementation plan to complete the corrective actions in a timely manner.
The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The intersystem leakage of the gas decay tanks increase the likelihood of a potential explosive mixture and continued to challenge technical specification oxygen concentration limits. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, because the finding is a contributor to a fire initiation event. The inspectors assigned a degradation rating of low to the finding since the oxygen concentration levels in the gas decay tanks were below the limit of an explosive mixture. The inspectors determined that the finding is of very low safety significance (Green) because the finding minimally impacted the fire protection capabilities of the fire area. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(2))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the FSAR following Modifications to the Reactor Coolant Pump Vapor Seals.
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) because the licensee did not revise the final safety analysis report (FSAR) as updated with information consistent with plant conditions.
Specifically, the licensee did not update Section 5.4.1.3 of the FSAR for Waterford Steam Electric Station, Unit 3 following modifications to the reactor coolant pump vapor seals in 2007 and 2009, respectively. As a result, the licensee did not promptly identify and correct FSAR noncompliance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7421. The planned corrective actions include revising the FSAR as updated and replacing the degraded reactor coolant pump seals during the next two refueling outages.
 
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because the longstanding and incorrect information in the FSAR as updated had a material impact on safety and licensed activities. The material impact is that the modifications created a reactor coolant pump seal loss of coolant accident likelihood inside containment, which could have potentially impacted licensed activities. The inspectors determined the violation is a Severity Level IV (very low safety significance) since the erroneous information not updated in the FSAR was not used to make an unacceptable change to the facility nor impacted a licensing or safety decision by the NRC. The inspectors determined there is a cross-cutting aspect in the corrective action component of the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate and take adequate actions in a timely manner to update the FSAR to be consistent with plant conditions [P.1.c of IMC 0310] (Section 1R18).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Work Order Instructions to Restore a Feedwater Heater Drain Valve.
A self-revealing finding occurred because maintenance personnel did not follow written procedures during the calibration of a level switch that controls feedwater heater drain valve FHD703A. Specifically, the licensee did not perform concurrent verification checks as required by documented work order instructions (WO-00180716) to ensure that personnel restore manipulate components to the correct position following maintenance. As a result, the feedwater heater drain valve remained in a closed manipulate state, which caused a spurious isolation of a string of feedwater heaters. The isolation of the feedwater heaters caused operators to down power the reactor to approximately 72 percent. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2009-7420.
The immediate corrective actions included restoring the feedwater heater drain valve to its proper position.
The finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the human error caused an event that upset plant stability during power operation. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it does not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensees personnel proceed in the face of uncertainty or unexpected circumstances [H.4.a of IMC 0310] (Section 4OA2.3).
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Determination Process for a Degraded and Non-Conforming condition Related to Reactor Coolant Pump N9000 Seals The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not adequately implement the operability determination process requirements in accordance with EN-OP-104, Operability Determination Process. Specifically, the licensee did not monitor a degraded and non-conformance condition associated with the reactor coolant pump N-9000 stage seals as required by EN-OP-104. As a result, the licensee did not perform a new operability determination after assumptions and compensatory measures identified in the original operability determination changed. This also led to compliance issues with technical specifications and missed maintenance rule functional failures. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1965. The immediate corrective actions included revising the operability determination to account for the current configuration. The planned corrective actions included the licensee replacing the degraded reactor coolant pump seals during the next two refueling outages.
 
The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition that had the potential to lead to a small loss of coolant accident. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification completion time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Perform Preventive Maintenance Activities Assiocated with the Dry Cooling Tower Process Analog Control Cards The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(3) because the licensee did not adequately evaluate and take into account, where practical, industry operating experience related to preventive maintenance activities for the dry cooling tower process analog control cards. Specifically, internal and industry-wide operating experience documented previous failures of process analog control cards due to age-related degradation after about 15 years of services. The vendor and industry performed a benchmark in 2003, and noted that the average service life is about 12 to 15 years. The licensee initially provided a preventive maintenance activity to replace the cards on a 20 year interval. However, the licensee deleted the preventive maintenance activities in March of 2009. The licensee determined that the cards were non-critical and had no justification of deleting the preventive maintenance activities.
The inspectors noted that after the deletion of the preventive maintenance activities and prior to the 15 year service internal, the licensee experienced additional unplanned failures of several process analog control cards that challenged dry cooling tower reliability. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1356. The immediate corrective action includes the evaluation of the preventive maintenance activity for the dry cooling tower process analog control cards. The planned corrective action includes the reinstatement of the preventive maintenance activity that aligns with industry operating experience.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The process analog control card failures challenged the system availability and reliability. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the condition is not a design or qualification deficiency, did not represent the loss of a system safety function, did not represent an actual loss of a single train of equipment for more than its Technical Specification completion time, and did not screen as potentially risk-significant due to an external initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through change to station processes, procedures, equipment, and training programs [P.2(b)]. (Section 1R12)
Inspection Report# : 2011004 (pdf)
 
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Work Order Instructions used for Technical Specification Surveillance Procedures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria.
The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)]. (Section 1R22.1)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Comply with Technical Specification Surveillance Requirement 4.0.3 and the Limiting Conditions for Operation for Technical Specifications 3.0.3 The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria.
The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and
 
Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)]. (Section 1R22.1)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Untimely Actions to Correct Repetitive Dry Cooling Tower Fan Failures The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(1))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit an LER within 60 days after Discovery of an Event The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations
 
Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(1))
Inspection Report# : 2011004 (pdf)
Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Monitor Activities Associated with the Internal Conditions of the Condensate and Refueling Water Storage Pool Structures.
The inspectors identified a non-cited violation of 10 CFR 50.65(a)(3) because the licensee did not evaluate or adequately monitor activities associated with the condition of the condensate and refueling water storage pools structures. Specifically, the licensee did not evaluate the internal condition of the storage pools through the performance of appropriate preventive maintenance activities and did not evaluate these activities at least every refueling cycle, where practical, for industry-wide operating experience. As a result, there is no preventive maintenance developed for this activity when previous industry-wide operating experience documented previous issues of concrete deterioration due to contact with boric acid over a long period of time. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1168. The planned corrective actions include the development of appropriate preventive maintenance activities to examine the internal conditions of the storage pool structures during the refuel outages.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, with no preventive maintenance to monitor the internal conditions of the storage pools, this would impact the reliability of the structures. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee did not implement and institutionalizes operating experience through changes to station processes, procedures, equipment, and training programs [P.2.b of IMC 0310] (Section 1R12).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Written Procedures for Restoring a Time Delay Relay Associated with the 'A' Emergency Diesel Generator Output Breaker.
A self-revealing non-cited violation of Technical Specification 6.8.1.a occurred because the licensee did not implement written procedures and instructions. Specifically, maintenance personnel did not follow procedure ME-007-005, Time Delay Relay Setting Check, Adjustment, and Functional Test, during the lifting leads process for restoration of a time delay relay (EG EREL2327-C) associated with the A emergency diesel generator (EDG) maintenance activity. As a result, the A EDG output breaker did not automatically close during technical specification surveillance testing because the leads on the relay were wired incorrectly. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3190. The immediate corrective action included the re-wiring of the relay.
The finding is more than minor because it is associated with the human and equipment performance attributes of the
 
Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the licensee did not ensure the availability, reliability and capability of the A EDG through human error prevention techniques. The senior resident inspector performed the initial significance determination for the diesel generator output breaker failure. The inspector used the NRC IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. The senior reactor analyst considered the output breaker a part of the emergency diesel generator component boundary. Assuming a one year exposure period, the finding was potentially Yellow, which warranted further review.
Therefore, the senior reactor analyst performed a bounding Phase 3 significance determination. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was approximately 5.4E-7/year. The dominant core damage sequences included loss of offsite power events, failure of the output breaker recovery action, independent failure of the other emergency diesel generator and failure to recover offsite power in 4 hours. Equipment that helped mitigate the risk included the ability of an operator to recover the output breaker. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not communicate human performance error prevention techniques, such as self and peer checking, and proper documentation of activities [H.4.a of IMC 0310] (Section 1R19).
Inspection Report# : 2011003 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Conduct Timely Corrective Actions to Replace Degraded Diodes in Safety Related Inverters A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, occurred because the licensee did not promptly correct a condition adverse to quality that affected static uninterruptible power supply inverters used to power vital and safety related loads. Specifically, the licensee did not conduct timely corrective actions following identification of degraded diodes in static uninterruptible power supplies A and B, respectively. As a result, this led to another failure of the static uninterruptible power supply A. The licensee entered this issue into their corrective action program (CAP) for resolution as CR-WF3-2010-6760. The immediate corrective actions following the additional failure included installation of newly tested diodes from a different batch, new fuses and a new silicon controlled rectifier. The planned corrective actions included implementation of an increased condition based testing preventive maintenance frequency and a maintenance activity to perform pre-installation testing on all new diodes and rectifiers.
This finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability and reliability of static uninterruptible power supply inverters that respond to initiating events to prevent undesirable consequences in that these inverters supply power to vital and safety related loads. The inspectors evaluated the significance of this finding using Phase 1 of the IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations given the importance of the system and the fact that this condition affects both static uninterruptible power supplies A and B. The inspectors determined that the finding was of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it Technical Specification allowed outage time, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the decision-making component of human performance because the licensee did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained.
Inspection Report# : 2010005 (pdf)
Barrier Integrity
 
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Control Room Envelope Preconditioning The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not conduct required technical specification surveillance testing on equipment in an as-found condition. Specifically, the licensee performed corrective maintenance (preconditioning) on the system to achieve better results, prior to completing the surveillance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1927. The immediate corrective action included the performance of the control room envelope tracer gas test.
The finding is more than minor because it is associated with the barrier performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not properly perform testing on equipment to evaluate barrier performance. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding doesnt represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room. The finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating the need for planned contingencies, compensatory actions, and abort criteria [H.3.a of IMC 0310] (Section 1R22).
Inspection Report# : 2011002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Aug 10, 2011 Identified By: NRC Item Type: FIN Finding Failure To Use Effective Engineering Controls As Part Of Pre-Job Planning To Reduce Contamination And Subsequent Exposure (Draft)
* TBD. The inspectors identified an apparent White finding because the licensee failed to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure. The primary reason for the dose overage was the licensees failure to prevent radioactive water from leaking into work areas and raising radiation dose rates. As corrective action, the licensee installed a trough system to collect and route the radioactive water away from the work area and to the reactor containment floor drain system. This issue was placed in the corrective action program as Condition Report CR-WF3-2011-05672.
The failure to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure is a performance deficiency. The finding is more than minor because it was similar to (the more than minor)
Example 6.i in Inspection Manual Chapter 0612, Appendix E, Example of Minor Issues, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Additionally, the finding is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective in that it increased collective radiation dose. The inspectors used Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to analyze the significance of the finding. The finding was preliminarily determined to be White (low to moderate safety significance) because it involved ALARA planning or work controls; the average collective dose at the time the
 
finding was identified was greater than 135 person-rem; and the actual dose associated with a work activity was greater than 25 person-rem. Alternately, there were greater than four occurrences in which the actual collective dose exceeded 5 person-rem and the estimated/planned dose by more than 50 percent. The final significance of this finding is to be determined. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with the operating experience component, because the licensee did not institutionalize operating experience concerning the effects of reactor coolant pump leakage on work area dose rates [P2.(b)] (Section 2RS02).
November 17, 2011, the NRC forwarded a letter that stated the final significance determination of a White inspection finding in the Occupational Radiation Safety Cornerstone (ML11321A291, EA2011-142). This letter provides the final significance determination of the preliminary White finding discussed in NRC Inspection Report 05000382/2011009, dated August 24, 2011. The finding involved the failure to use effective engineering controls to reduce radioactive contamination and subsequent exposure. Waterford Steam Electric Station personnel failed to keep highly radioactive water from leaking onto the work areas around the reactor coolant pumps. This failure resulted in high levels of radioactive contamination and unexpected and unintended radiation dose to plant workers.
Inspection Report# : 2011009 (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 Last modified : January 17, 2012
 
Waterford 3 4Q/2011 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement a Reactor Coolant System Drain Down Procedure The inspectors documented a self-revealing non-cited violation of Technical Specification 6.8.1.a because the licensee did not adequately implement Operating Procedure OP-001-003, Reactor Coolant System Drain Down, during the installation of the incore instrumentation flanges. Specifically, the licensee did not establish a reactor coolant system vent path while maintaining reactor coolant level below 26 feet for the assembly of the incore instrumentation flanges as required by OP-001-003. As a result, the licensee experienced a loss of reactor coolant inventory from three unassembled incore instrumentation flanges, which spilled onto the reactor vessel head insulation and filled the upper annulus cavity of the reactor vessel. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3163 and CR-WF3-2011-3636. The immediate corrective actions included opening the pressurizer spray line vent valve (RC-309) to establish a reactor coolant system vent path.
The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination for the failure to adequately implement operating procedures using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on the conditions of the plant at the time of the event. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately coordinate work activities in incorporating actions to address the impact of the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities
[H.3(b)]. (Section 1R20.1)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Testing for a Shutdown Cooling Heat Exchanger Outlet Stop Check Valve The inspectors documented a self-revealing non-cited violation of 10 CFR 50.55a, Codes and Standards, because the licensee did not establish and maintain an adequate testing program for a shutdown cooling heat exchanger outlet stop check valve (CS-117A) in accordance with Mandatory Appendix II, Check Valve Condition Monitoring Program, of the American Society of Mechanical Engineers Operation and Maintenance Code 2001 through 2003.
Specifically, the licensee did not provide adequate inservice testing to detect degradation of seat leakage on the stop check valve CS-117A. As a result, the operating train of shutdown cooling experienced a flow diversion when the licensee opened the upstream containment spray isolation header valve to fill the containment spray riser. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3350 and CR-WF3-2011-5841.
The immediate corrective action included the closure of the upstream isolation valve and the initiation of a work order to address seat leakage on the stop check valve CS-117. The planned corrective action includes the development of an augmented test to determine appropriate seat leakage criteria for the stop check valve.
 
The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 -
Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, since the degraded stop check valve upsets plant stability and challenge critical safety functions during shutdown conditions. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding did not have a cross-cutting aspect associated with it because the licensee established the check valve condition monitoring program prior to the past three years. Therefore it is not reflective of current plant performance. (Section 1R20.2)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Apparent Cause Evaluation Process Procedure The inspectors identified a finding because the licensee did not implement procedure EN-LI-119, Apparent Cause Evaluation Process. Specifically, the licensee did not follow the requirements provided in procedure EN-LI-119, Section 5.3.3 (k), to complete corrective actions in a timely manner for the intersystem leakage in the gas waste management system. As a result, no corrective action implementation occurred prior to additional equipment failures for the system. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0934. The immediate corrective action included the reevaluation of the causal determination and development of an implementation plan to complete the corrective actions in a timely manner.
The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The intersystem leakage of the gas decay tanks increase the likelihood of a potential explosive mixture and continued to challenge technical specification oxygen concentration limits. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, because the finding is a contributor to a fire initiation event. The inspectors assigned a degradation rating of low to the finding since the oxygen concentration levels in the gas decay tanks were below the limit of an explosive mixture. The inspectors determined that the finding is of very low safety significance (Green) because the finding minimally impacted the fire protection capabilities of the fire area. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(2))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the FSAR following Modifications to the Reactor Coolant Pump Vapor Seals.
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) because the licensee did not revise the final safety analysis report (FSAR) as updated with information consistent with plant conditions.
Specifically, the licensee did not update Section 5.4.1.3 of the FSAR for Waterford Steam Electric Station, Unit 3 following modifications to the reactor coolant pump vapor seals in 2007 and 2009, respectively. As a result, the licensee did not promptly identify and correct FSAR noncompliance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7421. The planned corrective actions include revising the FSAR as updated and replacing the degraded reactor coolant pump seals during the next two refueling outages.
 
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because the longstanding and incorrect information in the FSAR as updated had a material impact on safety and licensed activities. The material impact is that the modifications created a reactor coolant pump seal loss of coolant accident likelihood inside containment, which could have potentially impacted licensed activities. The inspectors determined the violation is a Severity Level IV (very low safety significance) since the erroneous information not updated in the FSAR was not used to make an unacceptable change to the facility nor impacted a licensing or safety decision by the NRC. The inspectors determined there is a cross-cutting aspect in the corrective action component of the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate and take adequate actions in a timely manner to update the FSAR to be consistent with plant conditions [P.1.c of IMC 0310] (Section 1R18).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Work Order Instructions to Restore a Feedwater Heater Drain Valve.
A self-revealing finding occurred because maintenance personnel did not follow written procedures during the calibration of a level switch that controls feedwater heater drain valve FHD703A. Specifically, the licensee did not perform concurrent verification checks as required by documented work order instructions (WO-00180716) to ensure that personnel restore manipulate components to the correct position following maintenance. As a result, the feedwater heater drain valve remained in a closed manipulate state, which caused a spurious isolation of a string of feedwater heaters. The isolation of the feedwater heaters caused operators to down power the reactor to approximately 72 percent. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2009-7420.
The immediate corrective actions included restoring the feedwater heater drain valve to its proper position.
The finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the human error caused an event that upset plant stability during power operation. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it does not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensees personnel proceed in the face of uncertainty or unexpected circumstances [H.4.a of IMC 0310] (Section 4OA2.3).
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Determination Process for a Degraded and Non-Conforming condition Related to Reactor Coolant Pump N9000 Seals The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not adequately implement the operability determination process requirements in accordance with EN-OP-104, Operability Determination Process. Specifically, the licensee did not monitor a degraded and non-conformance condition associated with the reactor coolant pump N-9000 stage seals as required by EN-OP-104. As a result, the licensee did not perform a new operability determination after assumptions and compensatory measures identified in the original operability determination changed. This also led to compliance issues with technical specifications and missed maintenance rule functional failures. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1965. The immediate corrective actions included revising the operability determination to account for the current configuration. The planned corrective actions included the licensee replacing the degraded reactor coolant pump seals during the next two refueling outages.
 
The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition that had the potential to lead to a small loss of coolant accident. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than its technical specification completion time, and did not screen as potentially risk significant due to external events. The finding has a cross-cutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Design Basis Input was Correctly Translated into Design Basis Calculations The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 28, 2011, the licensee failed to assure that design basis information associated with loading the auxiliary component cooling water pumps on the Class 1E Bus was correctly translated in various design basis calculations.
This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2011-06737 and CR-WF3-2011-06808.
The team determined that the failure to verify the adequacy of the design for loading the auxiliary component cooling water pumps on the Class 1E Bus in various design basis calculations was a performance deficiency. 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. Specifically, the inadequate design calculations could have prevented continued operation of the emergency diesel generator under degraded voltage, short circuit, and increased fuel oil consumption conditions. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee revised the associated calculations to include the required 295 brake horsepower value and reanalyzed for verification of operability. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Containment Spray Pump Design Basis Verification Surveillance Test The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Specifically, as of October 4,
 
2011, the licensee did not have an adequate test procedure to verify containment spray pump design basis accident performance requirements. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06852.
The team determined that the failure to either have a stand-alone design basis accident containment spray pump verification test or to have it adequately incorporated into the in-service testing requirements was a performance deficiency. 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. Specifically, neither the design basis analysis nor related in-service test surveillances, accounted for the inherent uncertainty of the flow element in the overall instrument uncertainty evaluation. In accordance with NRC Inspection Manual Chapter 0609, , "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Extrapolation of Vendor Supplied Pump Net Positive Suction Head Values The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of October 4, 2011, the licensee extrapolated the values for required pump net positive suction head beyond those provided in vendor certified curves without adequate analysis or justification. Consequently, the licensee, per the station-approved net positive suction head analysis, could have operated the safety-related pumps in beyond-analyzed or vendor-approved flow regimes. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06870.
The team determined that the failure to provide adequate justification for extrapolation of net positive suction head values beyond those provided in the certified pump vendor data was a performance deficiency. 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. Specifically, potential pump cavitation at higher than analyzed or vendor-approved operation, could have rendered mitigating equipment (i.e., pumps) to fail. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings,"
the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed additional analyses to assure that the pumps could safely operate in the required flow regimes. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Preventive Maintenance Procedures for Aluminum/Copper Electrical Connections to the Ultimate Heat Sink Transformers The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with
 
these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, as of October 7, 2011, when developing and implementing preventive maintenance procedures and work orders for transformers and electrical connections, the licensee failed to provide specific acceptance criteria and instructions addressing the potential vulnerability of these connections to degradation from galvanic reaction or differential thermal expansion, particularly in a high humidity outdoor environment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06832.
The team determined that the failure to provide suitable acceptance criteria and instructions in preventive maintenance procedures and work orders applicable to the aluminum/copper electrical connections to the transformers was a performance deficiency. 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. Specifically, inadequate preventive maintenance of the aluminum/copper connections could lead to degradation of the electrical connections to the station service transformer and loss of the ultimate heat sink. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis to Support the Adequacy of the Four Inch Bulkhead Drain to Protect the Ultimate Heat Sink During Flood Events The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event. Failure of the motor control center as a result of flooding from the probable maximum precipitation event could result in the loss of the associated ultimate heat sink because the motor control center serves both the dry cooling tower and wet cooling tower fan motors. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06701.
The team determined that the failure to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event was a performance deficiency. 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. Specifically, the design basis analysis for the four-inch bulkhead drain did not ensure that the motor control center would be adequately protected during a probable maximum precipitation event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed calculations to justify the adequacy of the installed bulkhead drain for the probable maximum precipitation event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
 
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis of the Effects of Reverse Rotation of Dry Cooling Tower Fan Motors Resulting from a Tornado Event The team identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to analyze the dry cooling tower fan motors for premature trip as a result of reverse rotation caused by a tornado event that could result in the loss of the dry cooling tower heat removal capability. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06850.
The team determined that the failure to establish and maintain an analysis supporting the ability of the dry cooling tower fan motors to operate successfully during and following a design basis tornado event was a performance deficiency. 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. Specifically, the design basis analysis did not ensure that the dry cooling tower fan motors would perform as required under reverse rotation conditions, without premature trip, during a design basis tornado. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee prepared an evaluation of the effect on fan motor starting current and duration for reverse rotation conditions. For reverse rotation conditions that would extend the locked rotor current time by a factor of two, the licensees analysis showed ample margin for the instantaneous trip settings from the magnetic-only breaker and the thermal overload protection, such that premature trip would be precluded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Temperature Limits of Auxiliary Component Cooling Water Pump Motor Bearings The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 7, 2011, the licensee did not have an adequate technical basis for increasing the auxiliary component cooling water pump motor bearing temperature alarm setpoints or establishing an upper limit on motor bearing temperature, which directed operators to secure the pump. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06573.
The team determined that the failure to provide an adequate basis for increasing the high bearing temperature alarm setpoints and establishing a high temperature motor trip criterion was a performance deficiency. 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. In accordance with NRC Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed an engineering justification for the bearing temperatures based on industry guidance. This finding was determined to have a cross-cutting aspect in the area of
 
human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage the Risk Involved with a Maintenance Window for the Turbine Driven Essential Feedwater Pump The team identified a Green noncited violation of 10 CFR 50.65(a)(4), which states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, on October 28, 2010 the turbine driven essential feedwater pump was out of service for maintenance for approximately 12 hours. During this time the licensee unknowingly entered the Orange risk window (crossed a risk threshold) due to a faulty assumption in the probabilistic risk assessment model. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06653.
The team determined that the failure to perform adequate risk assessments is a performance deficiency. This finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was identified as requiring a Phase 2 evaluation. A Region IV Senior Reactor Analyst performed a Phase 2 significance determination using NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K:
Delta-CDF = [CCDPActual - CCDPflawed]
* duration /8760 The licensee bounded the duration of the turbine driven essential feedwater pump maintenance at 8 hours in a year.
The flawed ICDP was 3.1E-5, the actual ICDP was 3.1E-5 + 1.9E-5 = 5.0E-5. The difference was 1.9E-5.
Delta-CDF = 1.9E-5
* 12/8760 = 2.6E-8 Therefore, the issue was determined to have very low safety significance (Green). This finding was determined to have a cross-cutting aspect in the area of problem identification and resolution associated with the self and independent assessments component because the licensee performed a probabilistic risk assessment model update in April 2009, which failed to identify the faulty assumption [P.3(a)].
Inspection Report# : 2011007 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Perform Preventive Maintenance Activities Assiocated with the Dry Cooling Tower Process Analog Control Cards The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(3) because the licensee did not adequately evaluate and take into account, where practical, industry operating experience related to preventive maintenance activities for the dry cooling tower process analog control cards. Specifically, internal and industry-wide operating experience documented previous failures of process analog control cards due to age-related degradation after about 15 years of services. The vendor and industry performed a benchmark in 2003, and noted that the average service life is about 12 to 15 years. The licensee initially provided a preventive maintenance activity to replace the cards on a 20 year interval. However, the licensee deleted the preventive maintenance activities in March of 2009. The licensee determined that the cards were non-critical and had no justification of deleting the preventive maintenance activities.
 
The inspectors noted that after the deletion of the preventive maintenance activities and prior to the 15 year service internal, the licensee experienced additional unplanned failures of several process analog control cards that challenged dry cooling tower reliability. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1356. The immediate corrective action includes the evaluation of the preventive maintenance activity for the dry cooling tower process analog control cards. The planned corrective action includes the reinstatement of the preventive maintenance activity that aligns with industry operating experience.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The process analog control card failures challenged the system availability and reliability. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the condition is not a design or qualification deficiency, did not represent the loss of a system safety function, did not represent an actual loss of a single train of equipment for more than its Technical Specification completion time, and did not screen as potentially risk-significant due to an external initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through change to station processes, procedures, equipment, and training programs [P.2(b)]. (Section 1R12)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Work Order Instructions used for Technical Specification Surveillance Procedures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria.
The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)]. (Section 1R22.1)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Comply with Technical Specification Surveillance Requirement 4.0.3 and the Limiting Conditions for Operation for Technical Specifications 3.0.3 The inspectors identified a non-cited violation of Technical Specification (TS) because the licensee did not enter or comply with the technical specification action requirements. Specifically, the licensee did not enter or comply with Technical Specification Surveillance Requirement 4.0.3 upon discovery of a never performed surveillance related to a safety-related relay contact for the Essential Chilled Water system. The licensee discovered the issue on July 27, 2011.
However, the licensee did not enter TS 4.0.3 until August 12, 2011. Subsequently, when the licensee entered TS 4.0.3, the licensee did not perform a risk evaluation within 24 hours, as directed by the technical specification surveillance requirement. The licensee, per Technical Specification 4.0.3, has up to 24 hours to perform a risk evaluation or enter the applicable technical specification limiting condition for operation immediately. The inspectors determined that the licensee exceeded the allowed 24 hours and then did not enter the limiting condition for operation for Technical Specification 3.0.3 once the requirements for Technical Specification 4.0.3 and other applicable technical specifications had not been met. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-5779. The immediate corrective action included the performance of a special test instruction to demonstrate operability of the safety-related relay.
The finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that a failure to comply with TS 4.0.3 and 3.0.3 affects the availability and reliability of the Essential Chill Water system. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, 609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in decision-making component of the human performance area because the licensee did not make a safety-significant or risk-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained [H.1(a)]. (Section 1R22.2)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Untimely Actions to Correct Repetitive Dry Cooling Tower Fan Failures The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource
 
component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(1))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit an LER within 60 days after Discovery of an Event The inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit required Licensee Event Reports (LERs) within 60 days after discovery of conditions that required a report. Specifically, the inspectors identified three instances of untimely LERs submittals for conditions related to an inoperable emergency feedwater pump, a single point vulnerability of spent fuel pool pumps, and a degraded fuel oil supply line for the Train A emergency diesel generator. The licensee submitted the reports at 332,163, and 101 days after discovery of the conditions, respectively. As a result, the licensee exceeded the 60 days for each condition that required a report.
The inspectors noted that this is also contrary to the licensees reportability procedure UNT-006-010, Event Notification and Reporting. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-5923. The immediate corrective actions include the performance of a human performance error review.
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because it occurred repeatedly within a two year period and the licensee missed opportunities to identify the issue.
The NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, this impacts the NRCs ability to carry out its statutory mission. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not define and effectively communicate expectations regarding procedural compliance
[H.4.(b)]. (Section 4OA3.4)
Inspection Report# : 2011004 (pdf)
Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Monitor Activities Associated with the Internal Conditions of the Condensate and Refueling Water Storage Pool Structures.
The inspectors identified a non-cited violation of 10 CFR 50.65(a)(3) because the licensee did not evaluate or adequately monitor activities associated with the condition of the condensate and refueling water storage pools structures. Specifically, the licensee did not evaluate the internal condition of the storage pools through the performance of appropriate preventive maintenance activities and did not evaluate these activities at least every refueling cycle, where practical, for industry-wide operating experience. As a result, there is no preventive maintenance developed for this activity when previous industry-wide operating experience documented previous issues of concrete deterioration due to contact with boric acid over a long period of time. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1168. The planned corrective actions include the development of appropriate preventive maintenance activities to examine the internal conditions of the storage pool structures during the refuel outages.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, with no preventive maintenance to monitor the internal conditions of the storage pools, this would impact the reliability of the structures. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in the operating experience component of the problem
 
identification and resolution area because the licensee did not implement and institutionalizes operating experience through changes to station processes, procedures, equipment, and training programs [P.2.b of IMC 0310] (Section 1R12).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Written Procedures for Restoring a Time Delay Relay Associated with the 'A' Emergency Diesel Generator Output Breaker.
A self-revealing non-cited violation of Technical Specification 6.8.1.a occurred because the licensee did not implement written procedures and instructions. Specifically, maintenance personnel did not follow procedure ME-007-005, Time Delay Relay Setting Check, Adjustment, and Functional Test, during the lifting leads process for restoration of a time delay relay (EG EREL2327-C) associated with the A emergency diesel generator (EDG) maintenance activity. As a result, the A EDG output breaker did not automatically close during technical specification surveillance testing because the leads on the relay were wired incorrectly. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3190. The immediate corrective action included the re-wiring of the relay.
The finding is more than minor because it is associated with the human and equipment performance attributes of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the licensee did not ensure the availability, reliability and capability of the A EDG through human error prevention techniques. The senior resident inspector performed the initial significance determination for the diesel generator output breaker failure. The inspector used the NRC IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. The senior reactor analyst considered the output breaker a part of the emergency diesel generator component boundary. Assuming a one year exposure period, the finding was potentially Yellow, which warranted further review.
Therefore, the senior reactor analyst performed a bounding Phase 3 significance determination. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was approximately 5.4E-7/year. The dominant core damage sequences included loss of offsite power events, failure of the output breaker recovery action, independent failure of the other emergency diesel generator and failure to recover offsite power in 4 hours. Equipment that helped mitigate the risk included the ability of an operator to recover the output breaker. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not communicate human performance error prevention techniques, such as self and peer checking, and proper documentation of activities [H.4.a of IMC 0310] (Section 1R19).
Inspection Report# : 2011003 (pdf)
Barrier Integrity Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Control Room Envelope Preconditioning The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not conduct required technical specification surveillance testing on equipment in an as-found condition. Specifically, the licensee performed corrective maintenance (preconditioning) on the system to achieve better results, prior to completing the surveillance. The licensee entered this issue into their corrective action program
 
for resolution as CR-WF3-2011-1927. The immediate corrective action included the performance of the control room envelope tracer gas test.
The finding is more than minor because it is associated with the barrier performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not properly perform testing on equipment to evaluate barrier performance. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding doesnt represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room. The finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating the need for planned contingencies, compensatory actions, and abort criteria [H.3.a of IMC 0310] (Section 1R22).
Inspection Report# : 2011002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Aug 10, 2011 Identified By: NRC Item Type: FIN Finding Failure To Use Effective Engineering Controls As Part Of Pre-Job Planning To Reduce Contamination And Subsequent Exposure The inspectors identified an apparent White finding because the licensee failed to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure. The primary reason for the dose overage was the licensees failure to prevent radioactive water from leaking into work areas and raising radiation dose rates.
As corrective action, the licensee installed a trough system to collect and route the radioactive water away from the work area and to the reactor containment floor drain system. This issue was placed in the corrective action program as Condition Report CR-WF3-2011-05672.
The failure to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure is a performance deficiency. The finding is more than minor because it was similar to (the more than minor)
Example 6.i in Inspection Manual Chapter 0612, Appendix E, Example of Minor Issues, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Additionally, the finding is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective in that it increased collective radiation dose. The inspectors used Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to analyze the significance of the finding. The finding was preliminarily determined to be White (low to moderate safety significance) because it involved ALARA planning or work controls; the average collective dose at the time the finding was identified was greater than 135 person-rem; and the actual dose associated with a work activity was greater than 25 person-rem. Alternately, there were greater than four occurrences in which the actual collective dose exceeded 5 person-rem and the estimated/planned dose by more than 50 percent. The final significance of this finding is to be determined. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with the operating experience component, because the licensee did not institutionalize operating experience concerning the effects of reactor coolant pump leakage on work area dose rates [P2.(b)] (Section 2RS02).
Inspection Report# : 2011009 (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 Last modified : March 02, 2012
 
Waterford 3 1Q/2012 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement a Reactor Coolant System Drain Down Procedure The inspectors documented a self-revealing non-cited violation of Technical Specification 6.8.1.a because the licensee did not adequately implement Operating Procedure OP-001-003, Reactor Coolant System Drain Down, during the installation of the incore instrumentation flanges. Specifically, the licensee did not establish a reactor coolant system vent path while maintaining reactor coolant level below 26 feet for the assembly of the incore instrumentation flanges as required by OP-001-003. As a result, the licensee experienced a loss of reactor coolant inventory from three unassembled incore instrumentation flanges, which spilled onto the reactor vessel head insulation and filled the upper annulus cavity of the reactor vessel. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3163 and CR-WF3-2011-3636. The immediate corrective actions included opening the pressurizer spray line vent valve (RC-309) to establish a reactor coolant system vent path.
The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination for the failure to adequately implement operating procedures using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on the conditions of the plant at the time of the event. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately coordinate work activities in incorporating actions to address the impact of the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities
[H.3(b)]. (Section 1R20.1)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Testing for a Shutdown Cooling Heat Exchanger Outlet Stop Check Valve The inspectors documented a self-revealing non-cited violation of 10 CFR 50.55a, Codes and Standards, because the licensee did not establish and maintain an adequate testing program for a shutdown cooling heat exchanger outlet stop check valve (CS-117A) in accordance with Mandatory Appendix II, Check Valve Condition Monitoring Program, of the American Society of Mechanical Engineers Operation and Maintenance Code 2001 through 2003.
Specifically, the licensee did not provide adequate inservice testing to detect degradation of seat leakage on the stop check valve CS-117A. As a result, the operating train of shutdown cooling experienced a flow diversion when the licensee opened the upstream containment spray isolation header valve to fill the containment spray riser. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3350 and CR-WF3-2011-5841.
The immediate corrective action included the closure of the upstream isolation valve and the initiation of a work order to address seat leakage on the stop check valve CS-117. The planned corrective action includes the development of an augmented test to determine appropriate seat leakage criteria for the stop check valve.
 
The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 -
Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, since the degraded stop check valve upsets plant stability and challenge critical safety functions during shutdown conditions. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding did not have a cross-cutting aspect associated with it because the licensee established the check valve condition monitoring program prior to the past three years. Therefore it is not reflective of current plant performance. (Section 1R20.2)
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Apparent Cause Evaluation Process Procedure The inspectors identified a finding because the licensee did not implement procedure EN-LI-119, Apparent Cause Evaluation Process. Specifically, the licensee did not follow the requirements provided in procedure EN-LI-119, Section 5.3.3 (k), to complete corrective actions in a timely manner for the intersystem leakage in the gas waste management system. As a result, no corrective action implementation occurred prior to additional equipment failures for the system. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0934. The immediate corrective action included the reevaluation of the causal determination and development of an implementation plan to complete the corrective actions in a timely manner.
The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The intersystem leakage of the gas decay tanks increase the likelihood of a potential explosive mixture and continued to challenge technical specification oxygen concentration limits. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, because the finding is a contributor to a fire initiation event. The inspectors assigned a degradation rating of low to the finding since the oxygen concentration levels in the gas decay tanks were below the limit of an explosive mixture. The inspectors determined that the finding is of very low safety significance (Green) because the finding minimally impacted the fire protection capabilities of the fire area. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(2))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the FSAR following Modifications to the Reactor Coolant Pump Vapor Seals.
The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) because the licensee did not revise the final safety analysis report (FSAR) as updated with information consistent with plant conditions.
Specifically, the licensee did not update Section 5.4.1.3 of the FSAR for Waterford Steam Electric Station, Unit 3 following modifications to the reactor coolant pump vapor seals in 2007 and 2009, respectively. As a result, the licensee did not promptly identify and correct FSAR noncompliance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7421. The planned corrective actions include revising the FSAR as updated and replacing the degraded reactor coolant pump seals during the next two refueling outages.
 
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because the longstanding and incorrect information in the FSAR as updated had a material impact on safety and licensed activities. The material impact is that the modifications created a reactor coolant pump seal loss of coolant accident likelihood inside containment, which could have potentially impacted licensed activities. The inspectors determined the violation is a Severity Level IV (very low safety significance) since the erroneous information not updated in the FSAR was not used to make an unacceptable change to the facility nor impacted a licensing or safety decision by the NRC. The inspectors determined there is a cross-cutting aspect in the corrective action component of the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate and take adequate actions in a timely manner to update the FSAR to be consistent with plant conditions [P.1.c of IMC 0310] (Section 1R18).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Work Order Instructions to Restore a Feedwater Heater Drain Valve.
A self-revealing finding occurred because maintenance personnel did not follow written procedures during the calibration of a level switch that controls feedwater heater drain valve FHD703A. Specifically, the licensee did not perform concurrent verification checks as required by documented work order instructions (WO-00180716) to ensure that personnel restore manipulate components to the correct position following maintenance. As a result, the feedwater heater drain valve remained in a closed manipulate state, which caused a spurious isolation of a string of feedwater heaters. The isolation of the feedwater heaters caused operators to down power the reactor to approximately 72 percent. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2009-7420.
The immediate corrective actions included restoring the feedwater heater drain valve to its proper position.
The finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the human error caused an event that upset plant stability during power operation. The inspectors evaluated this finding using IMC 0609 , Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it does not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensees personnel proceed in the face of uncertainty or unexpected circumstances [H.4.a of IMC 0310] (Section 4OA2.3).
Inspection Report# : 2011003 (pdf)
Mitigating Systems Significance:        Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Develop Preventive Maintenance Tasks for Critical Limit Switches on Component Cooling Water Inlet Isolation Valves A Green self-revealing non-cited violation of Waterford Steam Electric Station, Unit 3 Technical Specification 6.8.1.a occurred because the licensee did not establish procedures for performing preventive maintenance tasks on the dry cooling tower (DCT) component cooling water inlet isolation valves A and B (CC-135) limit switches. Specifically, the licensee did not develop preventive maintenance tasks to lubricate or replace critical limit switches that provide a permissive for the operation of the DCT fans. As a result, on February 4, 2011, the limit switch on valve CC-135A failed to operate as designed and rendered an entire train of DCT fans inoperable. The licensee entered this condition
 
into their corrective action program as CR-WF3-2011-0679 for resolution. The immediate corrective action included the lubrication of the limit switch and the manual stroking of the valve to obtain free and smooth movement of the degraded equipment. The planned corrective actions include the development of a preventive maintenance task to lubricate and replace the limit switches on a scheduled frequency.
The failure to establish procedures for performing preventive maintenance tasks on the dry cooling tower (DCT) component cooling water inlet isolation valves A and B (CC-135) limit switches is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since there is no preventive maintenance task for lubrication and replacement of the equipment, the limit switches can become stuck and render an entire train of DCT fans inoperable. The inspectors evaluated the significance determination using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings.
The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. The inspectors also concluded that no cross-cutting aspect is applicable to this finding because the performance deficiency is not reflective of current performance.
Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Performance of Safety-Related Valves The inspectors identified a Green non-cited violation of 10CFR50, Appendix B, Criterion XI because the licensee did not identify and perform testing on a safety-related component to demonstrate that it would perform satisfactory in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform proper testing for the Essential Chiller Hot Gas Bypass Valves (RFR-106A, B, and C). As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactory in service without performing a test and operability evaluation. The licensee entered this condition into the corrective action program as CR-WF3-2012-0632 and CR-WF3-2012-0659. The immediate corrective action included testing the Hot Gas Bypass Valves (HGBVs) to demonstrate the proper performance of their safety function.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactory in service in accordance with requirements contained in applicable design documents is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the hot gas bypass valve closure is required for essential chiller operation to maintain the reactor in a safe shutdown condition.
The inspectors evaluated the significance determination using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date test procedures were available to demonstrate that equipment performance is adequate to assure nuclear safety.
Inspection Report# : 2012002 (pdf)
Significance:        Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate the Impact of Fire Damage on the Dry Cooling Tower Fans
 
The team identified a non-cited violation of License Condition 2.C.9 and Appendix R, Section III.G for the failure to adequately evaluate the impact of fire damage on the dry cooling tower fans. Specifically, the failure to adequately evaluate fire damage to the dry cooling tower fans did not ensure one train remained available to achieve and maintain hot shutdown conditions from the alternate shutdown panel. The licensee documented this deficiency in Condition Report 2012-00837.
The failure to adequately evaluate the impact of fire damage on the dry cooling tower fans was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Since this finding involved a control room abandonment issue, a senior reactor analyst performed a Phase 3 significance determination. The senior reactor analyst determined this finding had very low risk significance based upon a bounding analysis (Green). The dominant core damage sequences involved a fire initiating event, failure of both the component cooling water and auxiliary component cooling water systems, as well as an independent failure of the turbine driven auxiliary feedwater pump.
Equipment that helped to mitigate the significance included the unaffected offsite power system, the viable steam generators and the safety related auxiliary feedwater system. Because the original failure to evaluate the impact of fire damage on the dry cooling tower fans had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Significance:      Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate Adequate Cooling Provided to Diesel Generator B within Required Time The team identified a non-cited violation of License Condition 2.C.9 and the fire protection program for the failure to perform a post-fire safe shutdown analysis design calculation. Specifically, the team determined that the licensee had not calculated the time available to establish component cooling water to prevent damaging the emergency diesel generator when providing power to post fire safe shutdown components. The licensee documented this deficiency in Condition Report 2012 00818.
The failure to perform a design calculation evaluating the ability to remove heat based upon emergency diesel generator loading following a control room fire was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Using Appendix F, the team assigned this finding a low degradation rating because the system was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. Specifically, the component cooling water system could accommodate the heat in the jacket water system of a lightly loaded diesel generator. This finding screened as very low safety significance (Green) in the Phase 1 evaluation. Because the original failure to perform a design calculation had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct a Condition Adverse to Quality Associated with the Main Feedwater Isolation
 
Valves Green. The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion XVI because the licensee failed to identify and correct a condition adverse to quality associated with the main feedwater isolation valve.
Specifically, the licensee did not identify that varnish deposits were causing the main feedwater isolation valve to fail its inservice testing. As a result, corrective actions that were implemented did not address the adverse condition, leading to a subsequent test failure. The licensee entered this issue into their corrective action program as CRWF3-2011-2005 and CR-WF3-2011-8140. The corrective actions included the replacement of the actuator, a shortening of the replacement frequency of the fourway hydraulic valves to a 36 month interval, and an evaluation of the current methods of gathering and implementing operating experience.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the main feedwater isolation valve is credited for closure during a main feedwater line break. The inspectors performed the initial significance determination using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a.However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment. Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment.
Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves.
Accordingly, the contribution to core damage was much less than E-6. Therefore, the inspectors determined that this finding had very low safety significance (Green). This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not collect and evaluate relevant external operating experience to identify that other sites experienced similar failures of feedwater isolation valves due to varnish deposits on the interior surface.
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Tornado Impact on the Ultimate Heat Sink During a Refueling Outage The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion III because the licensee did not translate applicable regulatory requirements and the design basis into specifications and instructions.
Specifically, the licensee did not translate the design basis tornado event into a design calculation. This outage-specific calculation was referenced by operations as the basis to ensure that the number of dry cooling tower fans needed for decay heat removal remained available. As a result, additional analysis needed to be performed to verify that the ultimate heat sink would have been able to perform its design function had a design basis tornado occurred during refueling outage RF-17. The licensee entered this issue into their corrective action program as CRWF3-2011-6480. The immediate corrective actions taken to restore compliance included analysis of the condition and actions to ensure that future outage specific calculations include the tornado design basis event.
The performance deficiency is more than minor because it challenges the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the calculation was used
 
when the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process, and Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Operational Checklists." The issue was determined to have a very low safety significance (Green) because it did not require a quantitative assessment. Through calculation review, the inspectors concluded that this failure resulted in the potential to enter an unanalyzed condition. This finding had a crosscutting aspect in the resources component of the human performance area in that the licensee failed to incorporate accurate design information into instructions.
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Work Order Instructions to Install a Swagelok Fitting on a Main Feedwater Isolation Valve Tube Connection The inspectors identified a non-cited violation of Technical Specification 6.8.1.a because the licensee did not follow work order instructions to install a pressure gage in an air line used to measure and maintain pressure for the hydraulic accumulators that close the main feedwater isolation valve. Specifically, the licensee did not follow the instructions to assemble and tighten a Swagelok fitting according to the work order. As a result, the fitting failed, preventing the valve from being able to perform its safety-related function. The licensee entered this issue into their corrective action program as CR-WF3-2010-1166 and CRWF3-2011-7469. The immediate corrective actions included repairing the Swagelok fitting and completing an apparent cause evaluation to determine the nature of the fitting failure and failure to follow procedure.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspector performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment.
Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment. Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves. Accordingly, the contribution to core damage was much less than E-6. As a result, this finding had a very low safety significance (Green). This finding does not have a crosscutting aspect since it is not indicative of current plant performance.
Inspection Report# : 2011005 (pdf)
 
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Design Basis Input was Correctly Translated into Design Basis Calculations The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 28, 2011, the licensee failed to assure that design basis information associated with loading the auxiliary component cooling water pumps on the Class 1E Bus was correctly translated in various design basis calculations.
This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2011-06737 and CR-WF3-2011-06808.
The team determined that the failure to verify the adequacy of the design for loading the auxiliary component cooling water pumps on the Class 1E Bus in various design basis calculations was a performance deficiency. 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. Specifically, the inadequate design calculations could have prevented continued operation of the emergency diesel generator under degraded voltage, short circuit, and increased fuel oil consumption conditions. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee revised the associated calculations to include the required 295 brake horsepower value and reanalyzed for verification of operability. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Containment Spray Pump Design Basis Verification Surveillance Test The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Specifically, as of October 4, 2011, the licensee did not have an adequate test procedure to verify containment spray pump design basis accident performance requirements. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06852.
The team determined that the failure to either have a stand-alone design basis accident containment spray pump verification test or to have it adequately incorporated into the in-service testing requirements was a performance deficiency. 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. Specifically, neither the design basis analysis nor related in-service test surveillances, accounted for the inherent uncertainty of the flow element in the overall instrument uncertainty evaluation. In accordance with NRC Inspection Manual Chapter 0609, , "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
 
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Extrapolation of Vendor Supplied Pump Net Positive Suction Head Values The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of October 4, 2011, the licensee extrapolated the values for required pump net positive suction head beyond those provided in vendor certified curves without adequate analysis or justification. Consequently, the licensee, per the station-approved net positive suction head analysis, could have operated the safety-related pumps in beyond-analyzed or vendor-approved flow regimes. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06870.
The team determined that the failure to provide adequate justification for extrapolation of net positive suction head values beyond those provided in the certified pump vendor data was a performance deficiency. 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. Specifically, potential pump cavitation at higher than analyzed or vendor-approved operation, could have rendered mitigating equipment (i.e., pumps) to fail. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings,"
the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed additional analyses to assure that the pumps could safely operate in the required flow regimes. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Preventive Maintenance Procedures for Aluminum/Copper Electrical Connections to the Ultimate Heat Sink Transformers The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, as of October 7, 2011, when developing and implementing preventive maintenance procedures and work orders for transformers and electrical connections, the licensee failed to provide specific acceptance criteria and instructions addressing the potential vulnerability of these connections to degradation from galvanic reaction or differential thermal expansion, particularly in a high humidity outdoor environment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06832.
The team determined that the failure to provide suitable acceptance criteria and instructions in preventive maintenance procedures and work orders applicable to the aluminum/copper electrical connections to the transformers was a performance deficiency. 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. Specifically, inadequate preventive maintenance of the aluminum/copper connections could lead to degradation of the electrical connections to the station service transformer and loss of the ultimate heat sink. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most
 
significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis to Support the Adequacy of the Four Inch Bulkhead Drain to Protect the Ultimate Heat Sink During Flood Events The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event. Failure of the motor control center as a result of flooding from the probable maximum precipitation event could result in the loss of the associated ultimate heat sink because the motor control center serves both the dry cooling tower and wet cooling tower fan motors. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06701.
The team determined that the failure to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event was a performance deficiency. 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. Specifically, the design basis analysis for the four-inch bulkhead drain did not ensure that the motor control center would be adequately protected during a probable maximum precipitation event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed calculations to justify the adequacy of the installed bulkhead drain for the probable maximum precipitation event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis of the Effects of Reverse Rotation of Dry Cooling Tower Fan Motors Resulting from a Tornado Event The team identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to analyze the dry cooling tower fan motors for premature trip as a result of reverse rotation caused by a tornado event that could result in the loss of the dry cooling tower heat removal capability. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06850.
The team determined that the failure to establish and maintain an analysis supporting the ability of the dry cooling tower fan motors to operate successfully during and following a design basis tornado event was a performance deficiency. 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. Specifically, the design basis analysis did not ensure that the dry cooling tower fan motors would perform as required under reverse rotation conditions, without premature trip, during a design basis tornado. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee prepared an evaluation of the effect on fan motor starting current and duration for reverse rotation conditions. For reverse rotation conditions that would extend the locked rotor current time by a factor of two, the licensees analysis showed ample margin for the instantaneous trip settings from the magnetic-only breaker and the thermal overload protection, such that premature trip would be precluded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Temperature Limits of Auxiliary Component Cooling Water Pump Motor Bearings The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 7, 2011, the licensee did not have an adequate technical basis for increasing the auxiliary component cooling water pump motor bearing temperature alarm setpoints or establishing an upper limit on motor bearing temperature, which directed operators to secure the pump. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06573.
The team determined that the failure to provide an adequate basis for increasing the high bearing temperature alarm setpoints and establishing a high temperature motor trip criterion was a performance deficiency. 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. In accordance with NRC Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed an engineering justification for the bearing temperatures based on industry guidance. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage the Risk Involved with a Maintenance Window for the Turbine Driven Essential Feedwater Pump The team identified a Green noncited violation of 10 CFR 50.65(a)(4), which states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, on October 28, 2010 the turbine driven essential feedwater pump was out of service for maintenance for approximately 12 hours. During this time the licensee unknowingly entered the Orange risk window (crossed a risk threshold) due to a faulty assumption in the probabilistic risk assessment model. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06653.
 
The team determined that the failure to perform adequate risk assessments is a performance deficiency. This finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was identified as requiring a Phase 2 evaluation. A Region IV Senior Reactor Analyst performed a Phase 2 significance determination using NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K:
Delta-CDF = [CCDPActual - CCDPflawed]
* duration /8760 The licensee bounded the duration of the turbine driven essential feedwater pump maintenance at 8 hours in a year.
The flawed ICDP was 3.1E-5, the actual ICDP was 3.1E-5 + 1.9E-5 = 5.0E-5. The difference was 1.9E-5.
Delta-CDF = 1.9E-5
* 12/8760 = 2.6E-8 Therefore, the issue was determined to have very low safety significance (Green). This finding was determined to have a cross-cutting aspect in the area of problem identification and resolution associated with the self and independent assessments component because the licensee performed a probabilistic risk assessment model update in April 2009, which failed to identify the faulty assumption [P.3(a)].
Inspection Report# : 2011007 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Perform Preventive Maintenance Activities Assiocated with the Dry Cooling Tower Process Analog Control Cards The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(3) because the licensee did not adequately evaluate and take into account, where practical, industry operating experience related to preventive maintenance activities for the dry cooling tower process analog control cards. Specifically, internal and industry-wide operating experience documented previous failures of process analog control cards due to age-related degradation after about 15 years of services. The vendor and industry performed a benchmark in 2003, and noted that the average service life is about 12 to 15 years. The licensee initially provided a preventive maintenance activity to replace the cards on a 20 year interval. However, the licensee deleted the preventive maintenance activities in March of 2009. The licensee determined that the cards were non-critical and had no justification of deleting the preventive maintenance activities.
The inspectors noted that after the deletion of the preventive maintenance activities and prior to the 15 year service internal, the licensee experienced additional unplanned failures of several process analog control cards that challenged dry cooling tower reliability. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1356. The immediate corrective action includes the evaluation of the preventive maintenance activity for the dry cooling tower process analog control cards. The planned corrective action includes the reinstatement of the preventive maintenance activity that aligns with industry operating experience.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The process analog control card failures challenged the system availability and reliability. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the condition is not a design or qualification deficiency, did not represent the loss of a system safety function, did not represent an actual loss of a single train of equipment for more than its Technical Specification completion time, and did not screen as potentially risk-significant due to an external initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through change to station processes, procedures, equipment, and training programs [P.2(b)]. (Section 1R12)
 
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Work Order Instructions used for Technical Specification Surveillance Procedures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria.
The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)]. (Section 1R22.1)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Comply with Technical Specification Surveillance Requirement 4.0.3 and the Limiting Conditions for Operation for Technical Specifications 3.0.3 The inspectors identified a non-cited violation of Technical Specification (TS) because the licensee did not enter or comply with the technical specification action requirements. Specifically, the licensee did not enter or comply with Technical Specification Surveillance Requirement 4.0.3 upon discovery of a never performed surveillance related to a safety-related relay contact for the Essential Chilled Water system. The licensee discovered the issue on July 27, 2011.
However, the licensee did not enter TS 4.0.3 until August 12, 2011. Subsequently, when the licensee entered TS 4.0.3, the licensee did not perform a risk evaluation within 24 hours, as directed by the technical specification surveillance requirement. The licensee, per Technical Specification 4.0.3, has up to 24 hours to perform a risk evaluation or enter the applicable technical specification limiting condition for operation immediately. The inspectors determined that the licensee exceeded the allowed 24 hours and then did not enter the limiting condition for operation for Technical Specification 3.0.3 once the requirements for Technical Specification 4.0.3 and other applicable technical specifications had not been met. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-5779. The immediate corrective action included the performance of a special test instruction to demonstrate operability of the safety-related relay.
The finding is more than minor because it is associated with the human performance attribute of the Mitigating
 
Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that a failure to comply with TS 4.0.3 and 3.0.3 affects the availability and reliability of the Essential Chill Water system. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, 609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in decision-making component of the human performance area because the licensee did not make a safety-significant or risk-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained [H.1(a)]. (Section 1R22.2)
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Untimely Actions to Correct Repetitive Dry Cooling Tower Fan Failures The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)]. (Section 4OA2.3(1))
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit an LER within 60 days after Discovery of an Event The inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit required Licensee Event Reports (LERs) within 60 days after discovery of conditions that required a report. Specifically, the inspectors identified three instances of untimely LERs submittals for conditions related to an inoperable emergency feedwater pump, a single point vulnerability of spent fuel pool pumps, and a degraded fuel oil supply line for the Train A emergency diesel generator. The licensee submitted the reports at 332,163, and 101 days after discovery of the conditions, respectively. As a result, the licensee exceeded the 60 days for each condition that required a report.
The inspectors noted that this is also contrary to the licensees reportability procedure UNT-006-010, Event Notification and Reporting. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-5923. The immediate corrective actions include the performance of a human performance error review.
 
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because it occurred repeatedly within a two year period and the licensee missed opportunities to identify the issue.
The NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, this impacts the NRCs ability to carry out its statutory mission. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not define and effectively communicate expectations regarding procedural compliance
[H.4.(b)]. (Section 4OA3.4)
Inspection Report# : 2011004 (pdf)
Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Monitor Activities Associated with the Internal Conditions of the Condensate and Refueling Water Storage Pool Structures.
The inspectors identified a non-cited violation of 10 CFR 50.65(a)(3) because the licensee did not evaluate or adequately monitor activities associated with the condition of the condensate and refueling water storage pools structures. Specifically, the licensee did not evaluate the internal condition of the storage pools through the performance of appropriate preventive maintenance activities and did not evaluate these activities at least every refueling cycle, where practical, for industry-wide operating experience. As a result, there is no preventive maintenance developed for this activity when previous industry-wide operating experience documented previous issues of concrete deterioration due to contact with boric acid over a long period of time. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1168. The planned corrective actions include the development of appropriate preventive maintenance activities to examine the internal conditions of the storage pool structures during the refuel outages.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, with no preventive maintenance to monitor the internal conditions of the storage pools, this would impact the reliability of the structures. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train greater than it technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee did not implement and institutionalizes operating experience through changes to station processes, procedures, equipment, and training programs [P.2.b of IMC 0310] (Section 1R12).
Inspection Report# : 2011003 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Written Procedures for Restoring a Time Delay Relay Associated with the 'A' Emergency Diesel Generator Output Breaker.
A self-revealing non-cited violation of Technical Specification 6.8.1.a occurred because the licensee did not implement written procedures and instructions. Specifically, maintenance personnel did not follow procedure ME-007-005, Time Delay Relay Setting Check, Adjustment, and Functional Test, during the lifting leads process for restoration of a time delay relay (EG EREL2327-C) associated with the A emergency diesel generator (EDG) maintenance activity. As a result, the A EDG output breaker did not automatically close during technical specification surveillance testing because the leads on the relay were wired incorrectly. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3190. The immediate corrective action included
 
the re-wiring of the relay.
The finding is more than minor because it is associated with the human and equipment performance attributes of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the licensee did not ensure the availability, reliability and capability of the A EDG through human error prevention techniques. The senior resident inspector performed the initial significance determination for the diesel generator output breaker failure. The inspector used the NRC IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. The senior reactor analyst considered the output breaker a part of the emergency diesel generator component boundary. Assuming a one year exposure period, the finding was potentially Yellow, which warranted further review.
Therefore, the senior reactor analyst performed a bounding Phase 3 significance determination. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was approximately 5.4E-7/year. The dominant core damage sequences included loss of offsite power events, failure of the output breaker recovery action, independent failure of the other emergency diesel generator and failure to recover offsite power in 4 hours. Equipment that helped mitigate the risk included the ability of an operator to recover the output breaker. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not communicate human performance error prevention techniques, such as self and peer checking, and proper documentation of activities [H.4.a of IMC 0310] (Section 1R19).
Inspection Report# : 2011003 (pdf)
Barrier Integrity Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inoperable Train of Containment Cooling System Green. The inspectors identified a non-cited violation of Technical Specification Limiting Condition for Operation 3.6.2.2, Containment Cooling System, which requires in Modes 1, 2, 3, and 4 that Two independent trains of containment cooling shall be OPERABLE with one fan cooler to each train. The Technical Specification Action statement requires that With one train of containment cooling inoperable, restore the inoperable train to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable containment cooling train to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours.
Specifically, from July 11, 2009, to July 19, 2009, the licensee failed to declare train B of the containment cooling system inoperable, and restore it to operable status within 72 hours or place the unit in hot standby in 6 hours. This finding has been entered into the licensees corrective action program as Condition Reports CR-WF3-2011-08150.
The inspectors determined that the failure to meet Technical Specification Limiting Condition for Operation 3.6.2.2 was a performance deficiency. The finding was more than minor because it adversely affected the structures, systems, and components and barrier performance attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the component cooling water flow for containment cooling system train B decreased below the minimum flow limits of Technical Specification Surveillance Requirement 4.6.2.2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment and heat removal components, and did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. This finding was determined to have a crosscutting aspect in the area of human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed
 
action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.
Inspection Report# : 2011005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Aug 10, 2011 Identified By: NRC Item Type: FIN Finding Failure To Use Effective Engineering Controls As Part Of Pre-Job Planning To Reduce Contamination And Subsequent Exposure The inspectors identified an apparent White finding because the licensee failed to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure. The primary reason for the dose overage was the licensees failure to prevent radioactive water from leaking into work areas and raising radiation dose rates.
As corrective action, the licensee installed a trough system to collect and route the radioactive water away from the work area and to the reactor containment floor drain system. This issue was placed in the corrective action program as Condition Report CR-WF3-2011-05672.
The failure to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure is a performance deficiency. The finding is more than minor because it was similar to (the more than minor)
Example 6.i in Inspection Manual Chapter 0612, Appendix E, Example of Minor Issues, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Additionally, the finding is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective in that it increased collective radiation dose. The inspectors used Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to analyze the significance of the finding. The finding was preliminarily determined to be White (low to moderate safety significance) because it involved ALARA planning or work controls; the average collective dose at the time the finding was identified was greater than 135 person-rem; and the actual dose associated with a work activity was greater than 25 person-rem. Alternately, there were greater than four occurrences in which the actual collective dose exceeded 5 person-rem and the estimated/planned dose by more than 50 percent. The final significance of this finding is to be determined. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with the operating experience component, because the licensee did not institutionalize operating experience concerning the effects of reactor coolant pump leakage on work area dose rates [P2.(b)] (Section 2RS02).
Inspection Report# : 2011009 (pdf)
Public Radiation Safety Significance: SL-IV Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Updated Final Safety Analysis Report The inspectors identified a Severity Level IV non-cited violation of 10 CFR Part 50.71, Maintenance of Records, because the licensee failed to update its Updated Final Safety Analysis Report (UFSAR) with submittals that include the effects of a change made to the facility. This finding was determined to be of very low safety significance.
 
While inspecting the licensees activities related to solid radwaste management and storage, the inspectors identified that the low level radwaste storage facility was not adequately described in Chapters 11 and 12 of the UFSAR. The licensee built the low level radwaste storage facility on the owner controlled area, outside of the protected area, for interim radwaste storage of dry active waste and solidified radioactive waste. Currently, the UFSAR, Chapters 11 and 12, Sections 11.4, Solid Waste Management, and 12.2.1, "Contained Radiation Sources," describe facilities for the storage of radioactive material, such as the dry active waste handling and spent resin handling system. Section 12.2.1.7 of the UFSAR also describes principal sources of radioactivity not enclosed by plant structures. This section included maximum activity inventory of different waste management system components, including the laundry tank, waste condensate tank, and spent resin tank. The low level radwaste storage facility was not described in the UFSAR in adequate detail. The licensee is committed to Regulatory Guide 1.70, Standard, Format, and Content of a Safety Analysis Report, Revision 2, dated September 1975, which describes the content of Chapter 11, Section 11.4, Solid Waste Management System. Regulatory Guide 1.70 states that this section should describe the capabilities of the plant to control, collect, handle, process, package, and temporarily store prior to shipment of solid radioactive waste generated as a result of normal operation, including anticipated operational occurrences. Regulatory Guide 1.70 also describes Chapter 12 of a safety analysis report and states, in part, that it should provide information on methods for radiation protection, estimated occupational radiation exposures to personnel during normal operation and anticipated operational occurrences, including radioactive material handling, processing, use, storage, and disposal. Section 12.2.1, Radiation Contained Sources, is the basis for the radiation protection design that should be described in the manner needed as input to the shield design calculations. Those sources that are contained in equipment like the radioactive waste management systems should be described. The source location in the plant should be specified so that all important sources of radioactivity can be located on plant layout drawings. Also, the UFSAR should provide a listing of isotope, quantity, form, and use of all sources that exceed 100 millicuries.
The low level radwaste storage facility has been in use since 1995 and contains a mixture of dry active waste and spent resin materials in separate storage compartments.
The 50.59 screening performed for this facility stated that the low level radwaste storage facility will have onsite storage space for a total of five years based on estimates of waste generation. This storage facility has been in operation for approximately 16 years.
The storage facility currently contains a significant source of radioactivity, 689.52 curies in total, which is not adequately described in the licensees UFSAR.
The performance deficiency associated with this finding was failure of the licensee to update the UFSAR to reflect changes made to the facility. This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that stored radwaste materials with a significant radioactive source term has been relocated from the plant radiologically controlled area to the owner controlled area. In addition, the radwaste management program has been affected because the licensee was not originally licensed to act as a low level waste facility. However, the termination of the Barnwell Low Level Radioactive Waste Management facility has forced the licensee to build such a storage area and make changes to the facility, significantly increasing the onsite storage capacity. The inspectors determined that this finding did not reflect present performance because it is an issue with changes made to the facility more than 15 years previously. Therefore, there was no cross-cutting aspect associated with this finding.
Inspection Report# : 2011005 (pdf)
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 : May 29, 2012
 
Waterford 3 2Q/2012 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to establish adequate procedural guidance to control feedwater heater level control valves A self-revealing finding occurred because the licensee did not establish adequate procedural guidance to control feedwater heater level control valves. Specifically, the procedures used to control the settings for the valves did not contain guidance that properly adjusted the proportional gain and air pressure input to ensure the valves open quickly during a transient. As a result, multiple failures in the feedwater heater drain system resulted in a feedwater pump A trip and a subsequent reactor power cutback. The licensee entered this condition into their corrective action program as CR-WF3-2012-1729 for resolution. The corrective actions included a revision of the procedure and loop calibration settings for the feedwater heater level control valves.
The failure to provide adequate guidance that properly adjusted the proportional gain to ensure the valves open as designed is a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, multiple feedwater heater control valve failures resulted in a reactor power cutback that upset plant stability. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date design documentation for loop calibration settings was available to assure nuclear safety [H.2(c)].
Inspection Report# : 2012003 (pdf)
Significance:        Sep 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adequately Implement a Reactor Coolant System Drain Down Procedure The inspectors documented a self-revealing non-cited violation of Technical Specification 6.8.1.a because the licensee did not adequately implement Operating Procedure OP-001-003, Reactor Coolant System Drain Down, during the installation of the incore instrumentation flanges. Specifically, the licensee did not establish a reactor coolant system vent path while maintaining reactor coolant level below 26 feet for the assembly of the incore instrumentation flanges as required by OP-001-003. As a result, the licensee experienced a loss of reactor coolant inventory from three unassembled incore instrumentation flanges, which spilled onto the reactor vessel head insulation and filled the upper annulus cavity of the reactor vessel. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3163 and CR-WF3-2011-3636. The immediate corrective actions included opening the pressurizer spray line vent valve (RC-309) to establish a reactor coolant system vent path.
The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination for the failure to adequately implement operating procedures using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance
 
Determination Process, based on the conditions of the plant at the time of the event. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately coordinate work activities in incorporating actions to address the impact of the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities
[H.3(b)].
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Testing for a Shutdown Cooling Heat Exchanger Outlet Stop Check Valve The inspectors documented a self-revealing non-cited violation of 10 CFR 50.55a, Codes and Standards, because the licensee did not establish and maintain an adequate testing program for a shutdown cooling heat exchanger outlet stop check valve (CS-117A) in accordance with Mandatory Appendix II, Check Valve Condition Monitoring Program, of the American Society of Mechanical Engineers Operation and Maintenance Code 2001 through 2003.
Specifically, the licensee did not provide adequate inservice testing to detect degradation of seat leakage on the stop check valve CS-117A. As a result, the operating train of shutdown cooling experienced a flow diversion when the licensee opened the upstream containment spray isolation header valve to fill the containment spray riser. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3350 and CR-WF3-2011-5841.
The immediate corrective action included the closure of the upstream isolation valve and the initiation of a work order to address seat leakage on the stop check valve CS-117. The planned corrective action includes the development of an augmented test to determine appropriate seat leakage criteria for the stop check valve.
The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 -
Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, since the degraded stop check valve upsets plant stability and challenge critical safety functions during shutdown conditions. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding did not have a cross-cutting aspect associated with it because the licensee established the check valve condition monitoring program prior to the past three years. Therefore it is not reflective of current plant performance.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Apparent Cause Evaluation Process Procedure The inspectors identified a finding because the licensee did not implement procedure EN-LI-119, Apparent Cause Evaluation Process. Specifically, the licensee did not follow the requirements provided in procedure EN-LI-119, Section 5.3.3 (k), to complete corrective actions in a timely manner for the intersystem leakage in the gas waste management system. As a result, no corrective action implementation occurred prior to additional equipment failures for the system. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0934. The immediate corrective action included the reevaluation of the causal determination and development of an implementation plan to complete the corrective actions in a timely manner.
The finding is more than minor because it is associated with the protection against external factors attribute of the
 
Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The intersystem leakage of the gas decay tanks increase the likelihood of a potential explosive mixture and continued to challenge technical specification oxygen concentration limits. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, because the finding is a contributor to a fire initiation event. The inspectors assigned a degradation rating of low to the finding since the oxygen concentration levels in the gas decay tanks were below the limit of an explosive mixture. The inspectors determined that the finding is of very low safety significance (Green) because the finding minimally impacted the fire protection capabilities of the fire area. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)].
Inspection Report# : 2011004 (pdf)
Mitigating Systems Significance:      Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III because the licensee did not provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis. Specifically, the licensee did not ensure that the design calculation used to determine the required number of wet cooling tower fans needed to operate the plant under normal and design conditions utilized the correct equation. As a result, the incorrect calculation provided reasonable doubt as to the operability of the wet cooling tower fans. The licensee entered this issue into their corrective action program as CR-WF3-2012-1395. The immediate corrective actions taken to restore compliance included a preliminary analysis of the condition and actions to perform a review of the methodology, inputs, and assumptions for the ultimate heat sink thermal performance calculations.
The failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the wet cooling tower fans are required to be operable for heat removal following all accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in a loss of operability or functionality of the ultimate heat sink. This finding has a cross-cutting aspect in the decision making component of the human performance area in that the licensee did not conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions [H.1(b)].
Inspection Report# : 2012003 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop Preventive Maintenance Tasks for Critical Limit Switches on Component Cooling Water Inlet Isolation Valves A Green self-revealing, non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification
 
6.8.1.a occurred because the licensee did not establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches. Specifically, the licensee had not developed preventive maintenance tasks to lubricate or replace critical limit switches that provide a permissive for the operation of the dry cooling tower fans. As a result, on February 4, 2011, the limit switch on valve CC-135A failed to operate as designed and rendered an entire train of fans inoperable. The licensee entered this condition into their corrective action program as CR-WF3-2011-0679 for resolution. The immediate corrective action included the lubrication of the limit switch and the manual stroking of the valve to obtain free and smooth movement of the degraded equipment. The planned corrective actions included the development of a preventive maintenance task to lubricate and replace the limit switches on a scheduled frequency.
The failure to establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since there was no preventive maintenance task for lubrication and replacement of the equipment, the limit switches can become stuck and render an entire train of dry cooling tower fans inoperable. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. The inspectors also concluded that no cross-cutting aspect is applicable to this finding because the performance deficiency is not reflective of current performance.
Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Performance of Safety-Related Valves The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, because the licensee did not identify and perform testing on a safety-related component to demonstrate that it would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform proper testing for the essential chiller hot gas bypass valves RFR-106A, B, and C. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service without performing a test and operability evaluation. The licensee entered this condition into the corrective action program as CR-WF3-2012-0632 and CR-WF3-2012-0659. The immediate corrective action included testing the hot gas bypass valves to demonstrate the proper performance of their safety function.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the hot gas bypass valve closure is required to ensure the essential chiller can perform its safety function during all design basis accident conditions. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to any external initiating events. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date test procedures were available to demonstrate that equipment performance is adequate to assure nuclear safety [H.2(c)]
Inspection Report# : 2012002 (pdf)
 
Significance:      Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate the Impact of Fire Damage on the Dry Cooling Tower Fans The team identified a non-cited violation of License Condition 2.C.9 and Appendix R, Section III.G for the failure to adequately evaluate the impact of fire damage on the dry cooling tower fans. Specifically, the failure to adequately evaluate fire damage to the dry cooling tower fans did not ensure one train remained available to achieve and maintain hot shutdown conditions from the alternate shutdown panel. The licensee documented this deficiency in Condition Report 2012-00837.
The failure to adequately evaluate the impact of fire damage on the dry cooling tower fans was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Since this finding involved a control room abandonment issue, a senior reactor analyst performed a Phase 3 significance determination. The senior reactor analyst determined this finding had very low risk significance based upon a bounding analysis (Green). The dominant core damage sequences involved a fire initiating event, failure of both the component cooling water and auxiliary component cooling water systems, as well as an independent failure of the turbine driven auxiliary feedwater pump.
Equipment that helped to mitigate the significance included the unaffected offsite power system, the viable steam generators and the safety related auxiliary feedwater system. Because the original failure to evaluate the impact of fire damage on the dry cooling tower fans had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Significance:      Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate Adequate Cooling Provided to Diesel Generator B within Required Time The team identified a non-cited violation of License Condition 2.C.9 and the fire protection program for the failure to perform a post-fire safe shutdown analysis design calculation. Specifically, the team determined that the licensee had not calculated the time available to establish component cooling water to prevent damaging the emergency diesel generator when providing power to post fire safe shutdown components. The licensee documented this deficiency in Condition Report 2012 00818.
The failure to perform a design calculation evaluating the ability to remove heat based upon emergency diesel generator loading following a control room fire was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Using Appendix F, the team assigned this finding a low degradation rating because the system was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. Specifically, the component cooling water system could accommodate the heat in the jacket water system of a lightly loaded diesel generator. This finding screened as very low safety significance (Green) in the Phase 1 evaluation. Because the original failure to perform a design calculation had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
 
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct a Condition Adverse to Quality Associated with the Main Feedwater Isolation Valves The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion XVI because the licensee failed to identify and correct a condition adverse to quality associated with the main feedwater isolation valve. Specifically, the licensee did not identify that varnish deposits were causing the main feedwater isolation valve to fail its inservice testing. As a result, corrective actions that were implemented did not address the adverse condition, leading to a subsequent test failure. The licensee entered this issue into their corrective action program as CRWF3-2011-2005 and CR-WF3-2011-8140. The corrective actions included the replacement of the actuator, a shortening of the replacement frequency of the fourway hydraulic valves to a 36 month interval, and an evaluation of the current methods of gathering and implementing operating experience.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the main feedwater isolation valve is credited for closure during a main feedwater line break. The inspectors performed the initial significance determination using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a.However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment. Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment.
Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves.
Accordingly, the contribution to core damage was much less than E-6. Therefore, the inspectors determined that this finding had very low safety significance (Green). This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not collect and evaluate relevant external operating experience to identify that other sites experienced similar failures of feedwater isolation valves due to varnish deposits on the interior surface [P.2(a)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Tornado Impact on the Ultimate Heat Sink During a Refueling Outage The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion III because the licensee did not translate applicable regulatory requirements and the design basis into specifications and instructions.
Specifically, the licensee did not translate the design basis tornado event into a design calculation. This outage-specific calculation was referenced by operations as the basis to ensure that the number of dry cooling tower fans needed for decay heat removal remained available. As a result, additional analysis needed to be performed to verify that the ultimate heat sink would have been able to perform its design function had a design basis tornado occurred during refueling outage RF-17. The licensee entered this issue into their corrective action program as CRWF3-
 
2011-6480. The immediate corrective actions taken to restore compliance included analysis of the condition and actions to ensure that future outage specific calculations include the tornado design basis event.
The performance deficiency is more than minor because it challenges the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the calculation was used when the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process, and Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Operational Checklists." The issue was determined to have a very low safety significance (Green) because it did not require a quantitative assessment. Through calculation review, the inspectors concluded that this failure resulted in the potential to enter an unanalyzed condition. This finding had a crosscutting aspect in the resources component of the human performance area in that the licensee failed to incorporate accurate design information into instructions [H.2(c)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Work Order Instructions to Install a Swagelok Fitting on a Main Feedwater Isolation Valve Tube Connection The inspectors identified a non-cited violation of Technical Specification 6.8.1.a because the licensee did not follow work order instructions to install a pressure gage in an air line used to measure and maintain pressure for the hydraulic accumulators that close the main feedwater isolation valve. Specifically, the licensee did not follow the instructions to assemble and tighten a Swagelok fitting according to the work order. As a result, the fitting failed, preventing the valve from being able to perform its safety-related function. The licensee entered this issue into their corrective action program as CR-WF3-2010-1166 and CRWF3-2011-7469. The immediate corrective actions included repairing the Swagelok fitting and completing an apparent cause evaluation to determine the nature of the fitting failure and failure to follow procedure.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspector performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment.
 
Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment. Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves. Accordingly, the contribution to core damage was much less than E-6. As a result, this finding had a very low safety significance (Green). This finding does not have a crosscutting aspect since it is not indicative of current plant performance.
Inspection Report# : 2011005 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Design Basis Input was Correctly Translated into Design Basis Calculations The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 28, 2011, the licensee failed to assure that design basis information associated with loading the auxiliary component cooling water pumps on the Class 1E Bus was correctly translated in various design basis calculations.
This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2011-06737 and CR-WF3-2011-06808.
The team determined that the failure to verify the adequacy of the design for loading the auxiliary component cooling water pumps on the Class 1E Bus in various design basis calculations was a performance deficiency. 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. Specifically, the inadequate design calculations could have prevented continued operation of the emergency diesel generator under degraded voltage, short circuit, and increased fuel oil consumption conditions. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee revised the associated calculations to include the required 295 brake horsepower value and reanalyzed for verification of operability. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Containment Spray Pump Design Basis Verification Surveillance Test The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Specifically, as of October 4, 2011, the licensee did not have an adequate test procedure to verify containment spray pump design basis accident performance requirements. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06852.
The team determined that the failure to either have a stand-alone design basis accident containment spray pump verification test or to have it adequately incorporated into the in-service testing requirements was a performance deficiency. 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. Specifically, neither the design basis analysis nor related in-service test surveillances, accounted for the inherent uncertainty of the flow element in the overall instrument uncertainty evaluation. In accordance with NRC Inspection Manual Chapter 0609, , "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Extrapolation of Vendor Supplied Pump Net Positive Suction Head Values The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of October 4, 2011, the licensee extrapolated the values for required pump net positive suction head beyond those provided in vendor certified curves without adequate analysis or justification. Consequently, the licensee, per the station-approved net positive suction head analysis, could have operated the safety-related pumps in beyond-analyzed or vendor-approved flow regimes. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06870.
The team determined that the failure to provide adequate justification for extrapolation of net positive suction head values beyond those provided in the certified pump vendor data was a performance deficiency. 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. Specifically, potential pump cavitation at higher than analyzed or vendor-approved operation, could have rendered mitigating equipment (i.e., pumps) to fail. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings,"
the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed additional analyses to assure that the pumps could safely operate in the required flow regimes. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Preventive Maintenance Procedures for Aluminum/Copper Electrical Connections to the Ultimate Heat Sink Transformers The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, as of October 7, 2011, when developing and implementing preventive maintenance procedures and work orders for transformers and electrical connections, the licensee failed to provide specific acceptance criteria and instructions addressing the potential vulnerability of these connections to degradation from galvanic reaction or differential thermal expansion, particularly in a high humidity outdoor environment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06832.
 
The team determined that the failure to provide suitable acceptance criteria and instructions in preventive maintenance procedures and work orders applicable to the aluminum/copper electrical connections to the transformers was a performance deficiency. 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. Specifically, inadequate preventive maintenance of the aluminum/copper connections could lead to degradation of the electrical connections to the station service transformer and loss of the ultimate heat sink. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis to Support the Adequacy of the Four Inch Bulkhead Drain to Protect the Ultimate Heat Sink During Flood Events The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event. Failure of the motor control center as a result of flooding from the probable maximum precipitation event could result in the loss of the associated ultimate heat sink because the motor control center serves both the dry cooling tower and wet cooling tower fan motors. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06701.
The team determined that the failure to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event was a performance deficiency. 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. Specifically, the design basis analysis for the four-inch bulkhead drain did not ensure that the motor control center would be adequately protected during a probable maximum precipitation event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed calculations to justify the adequacy of the installed bulkhead drain for the probable maximum precipitation event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis of the Effects of Reverse Rotation of Dry Cooling Tower Fan Motors Resulting from a Tornado Event
 
The team identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to analyze the dry cooling tower fan motors for premature trip as a result of reverse rotation caused by a tornado event that could result in the loss of the dry cooling tower heat removal capability. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06850.
The team determined that the failure to establish and maintain an analysis supporting the ability of the dry cooling tower fan motors to operate successfully during and following a design basis tornado event was a performance deficiency. 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. Specifically, the design basis analysis did not ensure that the dry cooling tower fan motors would perform as required under reverse rotation conditions, without premature trip, during a design basis tornado. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee prepared an evaluation of the effect on fan motor starting current and duration for reverse rotation conditions. For reverse rotation conditions that would extend the locked rotor current time by a factor of two, the licensees analysis showed ample margin for the instantaneous trip settings from the magnetic-only breaker and the thermal overload protection, such that premature trip would be precluded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Temperature Limits of Auxiliary Component Cooling Water Pump Motor Bearings The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 7, 2011, the licensee did not have an adequate technical basis for increasing the auxiliary component cooling water pump motor bearing temperature alarm setpoints or establishing an upper limit on motor bearing temperature, which directed operators to secure the pump. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06573.
The team determined that the failure to provide an adequate basis for increasing the high bearing temperature alarm setpoints and establishing a high temperature motor trip criterion was a performance deficiency. 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. In accordance with NRC Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed an engineering justification for the bearing temperatures based on industry guidance. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
 
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage the Risk Involved with a Maintenance Window for the Turbine Driven Essential Feedwater Pump The team identified a Green noncited violation of 10 CFR 50.65(a)(4), which states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, on October 28, 2010 the turbine driven essential feedwater pump was out of service for maintenance for approximately 12 hours. During this time the licensee unknowingly entered the Orange risk window (crossed a risk threshold) due to a faulty assumption in the probabilistic risk assessment model. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06653.
The team determined that the failure to perform adequate risk assessments is a performance deficiency. This finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was identified as requiring a Phase 2 evaluation. A Region IV Senior Reactor Analyst performed a Phase 2 significance determination using NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K:
Delta-CDF = [CCDPActual - CCDPflawed]
* duration /8760 The licensee bounded the duration of the turbine driven essential feedwater pump maintenance at 8 hours in a year.
The flawed ICDP was 3.1E-5, the actual ICDP was 3.1E-5 + 1.9E-5 = 5.0E-5. The difference was 1.9E-5.
Delta-CDF = 1.9E-5
* 12/8760 = 2.6E-8 Therefore, the issue was determined to have very low safety significance (Green). This finding was determined to have a cross-cutting aspect in the area of problem identification and resolution associated with the self and independent assessments component because the licensee performed a probabilistic risk assessment model update in April 2009, which failed to identify the faulty assumption [P.3(a)].
Inspection Report# : 2011007 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate and Adequately Perform Preventive Maintenance Activities Assiocated with the Dry Cooling Tower Process Analog Control Cards The inspectors identified a non-cited violation of 10 CFR 50.65 (a)(3) because the licensee did not adequately evaluate and take into account, where practical, industry operating experience related to preventive maintenance activities for the dry cooling tower process analog control cards. Specifically, internal and industry-wide operating experience documented previous failures of process analog control cards due to age-related degradation after about 15 years of services. The vendor and industry performed a benchmark in 2003, and noted that the average service life is about 12 to 15 years. The licensee initially provided a preventive maintenance activity to replace the cards on a 20 year interval. However, the licensee deleted the preventive maintenance activities in March of 2009. The licensee determined that the cards were non-critical and had no justification of deleting the preventive maintenance activities.
The inspectors noted that after the deletion of the preventive maintenance activities and prior to the 15 year service internal, the licensee experienced additional unplanned failures of several process analog control cards that challenged dry cooling tower reliability. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1356. The immediate corrective action includes the evaluation of the preventive maintenance activity for
 
the dry cooling tower process analog control cards. The planned corrective action includes the reinstatement of the preventive maintenance activity that aligns with industry operating experience.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The process analog control card failures challenged the system availability and reliability. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the condition is not a design or qualification deficiency, did not represent the loss of a system safety function, did not represent an actual loss of a single train of equipment for more than its Technical Specification completion time, and did not screen as potentially risk-significant due to an external initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through change to station processes, procedures, equipment, and training programs [P.2(b)].
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct Work Order Instructions used for Technical Specification Surveillance Procedures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria.
The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Comply with Technical Specification Surveillance Requirement 4.0.3 and the Limiting Conditions for Operation for Technical Specifications 3.0.3 The inspectors identified a non-cited violation of Technical Specification (TS) because the licensee did not enter or
 
comply with the technical specification action requirements. Specifically, the licensee did not enter or comply with Technical Specification Surveillance Requirement 4.0.3 upon discovery of a never performed surveillance related to a safety-related relay contact for the Essential Chilled Water system. The licensee discovered the issue on July 27, 2011.
However, the licensee did not enter TS 4.0.3 until August 12, 2011. Subsequently, when the licensee entered TS 4.0.3, the licensee did not perform a risk evaluation within 24 hours, as directed by the technical specification surveillance requirement. The licensee, per Technical Specification 4.0.3, has up to 24 hours to perform a risk evaluation or enter the applicable technical specification limiting condition for operation immediately. The inspectors determined that the licensee exceeded the allowed 24 hours and then did not enter the limiting condition for operation for Technical Specification 3.0.3 once the requirements for Technical Specification 4.0.3 and other applicable technical specifications had not been met. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-5779. The immediate corrective action included the performance of a special test instruction to demonstrate operability of the safety-related relay.
The finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that a failure to comply with TS 4.0.3 and 3.0.3 affects the availability and reliability of the Essential Chill Water system. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, 609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in decision-making component of the human performance area because the licensee did not make a safety-significant or risk-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained [H.1(a)].
Inspection Report# : 2011004 (pdf)
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Untimely Actions to Correct Repetitive Dry Cooling Tower Fan Failures The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design.
The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals [H.2(a)].
Inspection Report# : 2011004 (pdf)
Significance: SL-IV Sep 30, 2011
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit an LER within 60 days after Discovery of an Event The inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit required Licensee Event Reports (LERs) within 60 days after discovery of conditions that required a report. Specifically, the inspectors identified three instances of untimely LERs submittals for conditions related to an inoperable emergency feedwater pump, a single point vulnerability of spent fuel pool pumps, and a degraded fuel oil supply line for the Train A emergency diesel generator. The licensee submitted the reports at 332,163, and 101 days after discovery of the conditions, respectively. As a result, the licensee exceeded the 60 days for each condition that required a report.
The inspectors noted that this is also contrary to the licensees reportability procedure UNT-006-010, Event Notification and Reporting. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-5923. The immediate corrective actions include the performance of a human performance error review.
The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC's ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because it occurred repeatedly within a two year period and the licensee missed opportunities to identify the issue.
The NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, this impacts the NRCs ability to carry out its statutory mission. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not define and effectively communicate expectations regarding procedural compliance
[H.4.(b)].
Inspection Report# : 2011004 (pdf)
Barrier Integrity Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inoperable Train of Containment Cooling System The inspectors identified a non-cited violation of Technical Specification Limiting Condition for Operation 3.6.2.2, Containment Cooling System, which requires in Modes 1, 2, 3, and 4 that Two independent trains of containment cooling shall be OPERABLE with one fan cooler to each train. The Technical Specification Action statement requires that With one train of containment cooling inoperable, restore the inoperable train to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable containment cooling train to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours. Specifically, from July 11, 2009, to July 19, 2009, the licensee failed to declare train B of the containment cooling system inoperable, and restore it to operable status within 72 hours or place the unit in hot standby in 6 hours. This finding has been entered into the licensees corrective action program as Condition Reports CR-WF3-2011-08150.
The inspectors determined that the failure to meet Technical Specification Limiting Condition for Operation 3.6.2.2 was a performance deficiency. The finding was more than minor because it adversely affected the structures, systems, and components and barrier performance attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the component cooling water flow for containment cooling system train B decreased below the minimum flow limits of Technical Specification Surveillance Requirement 4.6.2.2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment and heat removal components, and did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. This finding was determined to have a crosscutting aspect in the area of human performance associated with the decision making component because the licensee
 
did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2011005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Aug 10, 2011 Identified By: NRC Item Type: FIN Finding Failure To Use Effective Engineering Controls As Part Of Pre-Job Planning To Reduce Contamination And Subsequent Exposure The inspectors identified an apparent White finding because the licensee failed to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure. The primary reason for the dose overage was the licensees failure to prevent radioactive water from leaking into work areas and raising radiation dose rates.
As corrective action, the licensee installed a trough system to collect and route the radioactive water away from the work area and to the reactor containment floor drain system. This issue was placed in the corrective action program as Condition Report CR-WF3-2011-05672.
The failure to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure is a performance deficiency. The finding is more than minor because it was similar to (the more than minor)
Example 6.i in Inspection Manual Chapter 0612, Appendix E, Example of Minor Issues, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Additionally, the finding is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective in that it increased collective radiation dose. The inspectors used Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to analyze the significance of the finding. The finding was preliminarily determined to be White (low to moderate safety significance) because it involved ALARA planning or work controls; the average collective dose at the time the finding was identified was greater than 135 person-rem; and the actual dose associated with a work activity was greater than 25 person-rem. Alternately, there were greater than four occurrences in which the actual collective dose exceeded 5 person-rem and the estimated/planned dose by more than 50 percent. The final significance of this finding is to be determined. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with the operating experience component, because the licensee did not institutionalize operating experience concerning the effects of reactor coolant pump leakage on work area dose rates [P2.(b)] (Section 2RS02).
UPDATE -
In spite of the known fuel defects and planned crud burst with anticipated reactor coolant seal leakage, Waterford Steam Electric Station personnel failed to install an effective leak collection system to collect and control the seal leakage. The resulting contamination of work areas caused unnecessary personnel dose well in excess of the original ALARA dose estimates.
Based on this, the NRC has concluded that the finding is appropriately characterized as White (i.e., a finding with a low-to-moderate importance to safety that may require additional NRC inspections). Using Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the finding was determined to be White because (1) this was an ALARA planning issue, (2) the site's 3-year average collective dose exceeded 135 person-rem and (3) one of the work activities accrued more than 25 person-rem or alternatively, the finding would still be White because there were more than four other occurrences in which the actual collective dose exceeded 5 person-rem, and the estimated/planned dose by more than 50 percent.
 
(Letter "Waterford Steam Electric Station, Unit 3 -Final Significance Determination of White Finding, NRC Inspection Report 05000382/2011009, dated November 17, 2011; ML11321A291)
Update -
The NRC staff performed this supplemental inspection in accordance with IP 95001, Inspection for One or Two White Inputs in a Strategic Performance Area, to assess the licensee=s evaluation associated with the failure to use effective engineering controls as part of prejob planning to reduce contamination and subsequent exposure.
The NRC staff previously characterized this issue as having low to moderate safety significance (White), as documented in NRC IR 05000382/2011009. During this supplemental inspection, the inspector determined the licensee performed a comprehensive evaluation of the NRC-identified failure, which occurred during Refueling Outage 16. The licensee identified the primary root cause of the issue was response to the known, repetitive leakage was untimely because personnel exhibited insufficient awareness of the impact of actions on safety and reliability.
Contributing causes were also identified. The licensees review of the extent of cause revealed the root cause was not limited to the use of leakage control devices. Additionally, the licensees staff members may not have completely understood the level of radiological risk associated with floor drains with poor drainage. The licensee has taken corrective actions to address the root cause by installing an engineered drain system to route contaminated water leaking from reactor coolant pump seals and by implementing procedural controls that will require the review and evaluation of contaminated leakage control devices for known radiologically significant leakage problems.
Given the licensees acceptable performance in addressing the leaking reactor coolant, the (white) finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Inspectors will review the licensees implementation of corrective actions during a future inspection.
(IR 05000382/2012010 dated July 5, 2012, ML12187A725)
Inspection Report# : 2011009 (pdf)
Inspection Report# : 2012010 (pdf)
Public Radiation Safety Significance: SL-IV Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Updated Final Safety Analysis Report The inspectors identified a non-cited violation of 10 CFR 50.71 Maintenance of Records, because the licensee failed to update their updated final safety analysis report with submittals that include a change made to the facility.
Specifically, the licensee built the low level radwaste storage facility in 1995 on the owner controlled area for interim radwaste storage of dry and solidified radioactive waste and failed to update the updated final safety analysis report to include these changes. This issue was entered in the licensees corrective action program as condition report WF3-2011-07711.
This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that stored radwaste materials with a significant radioactive source term has been relocated from the plant radiologically controlled area to the owner controlled area.
In addition, the radwaste management program has been affected because the licensee was not originally licensed to act as a low level waste facility. However, the termination of the Barnwell Low Level Radioactive Waste Management facility has forced the licensee to build such a storage area and make changes to the facility, significantly increasing the onsite storage capacity. The inspectors determined that this finding did not reflect present performance
 
because it is an issue with changes made to the facility more than 15 years previously. Therefore, there was no cross-cutting aspect associated with this finding. This finding is characterized as a Severity Level IV non-cited violation in accordance with NRC Enforcement Policy, Section 6.1, and was treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy Inspection Report# : 2011005 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : September 12, 2012
 
3Q/2012 Inspection Findings - Waterford 3 Waterford 3 3Q/2012 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to establish adequate procedural guidance to control feedwater heater level control valves A self-revealing finding occurred because the licensee did not establish adequate procedural guidance to control feedwater heater level control valves. Specifically, the procedures used to control the settings for the valves did not contain guidance that properly adjusted the proportional gain and air pressure input to ensure the valves open quickly during a transient. As a result, multiple failures in the feedwater heater drain system resulted in a feedwater pump A trip and a subsequent reactor power cutback. The licensee entered this condition into their corrective action program as CR-WF3-2012-1729 for resolution. The corrective actions included a revision of the procedure and loop calibration settings for the feedwater heater level control valves.
The failure to provide adequate guidance that properly adjusted the proportional gain to ensure the valves open as designed is a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, multiple feedwater heater control valve failures resulted in a reactor power cutback that upset plant stability. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date design documentation for loop calibration settings was available to assure nuclear safety [H.2(c)].
Inspection Report# : 2012003 (pdf)
Mitigating Systems Significance:      Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III because the licensee did not provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis. Specifically, the licensee did not ensure that the design calculation used to determine the required number of wet cooling tower fans needed to operate the plant under normal and design conditions utilized the correct equation. As a result, the incorrect calculation provided reasonable doubt as to the operability of the wet Page 1 of 13
 
3Q/2012 Inspection Findings - Waterford 3 cooling tower fans. The licensee entered this issue into their corrective action program as CR-WF3-2012-1395. The immediate corrective actions taken to restore compliance included a preliminary analysis of the condition and actions to perform a review of the methodology, inputs, and assumptions for the ultimate heat sink thermal performance calculations.
The failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the wet cooling tower fans are required to be operable for heat removal following all accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in a loss of operability or functionality of the ultimate heat sink. This finding has a cross-cutting aspect in the decision making component of the human performance area in that the licensee did not conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions [H.1(b)].
Inspection Report# : 2012003 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop Preventive Maintenance Tasks for Critical Limit Switches on Component Cooling Water Inlet Isolation Valves A Green self-revealing, non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a occurred because the licensee did not establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches. Specifically, the licensee had not developed preventive maintenance tasks to lubricate or replace critical limit switches that provide a permissive for the operation of the dry cooling tower fans. As a result, on February 4, 2011, the limit switch on valve CC-135A failed to operate as designed and rendered an entire train of fans inoperable. The licensee entered this condition into their corrective action program as CR-WF3-2011-0679 for resolution. The immediate corrective action included the lubrication of the limit switch and the manual stroking of the valve to obtain free and smooth movement of the degraded equipment. The planned corrective actions included the development of a preventive maintenance task to lubricate and replace the limit switches on a scheduled frequency.
The failure to establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since there was no preventive maintenance task for lubrication and replacement of the equipment, the limit switches can become stuck and render an entire train of dry cooling tower fans inoperable. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. The inspectors also concluded that no cross-cutting aspect is applicable to this finding because the performance deficiency is not reflective of current performance.
Page 2 of 13
 
3Q/2012 Inspection Findings - Waterford 3 Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Performance of Safety-Related Valves The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, because the licensee did not identify and perform testing on a safety-related component to demonstrate that it would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform proper testing for the essential chiller hot gas bypass valves RFR-106A, B, and C. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service without performing a test and operability evaluation. The licensee entered this condition into the corrective action program as CR-WF3-2012-0632 and CR-WF3-2012-0659. The immediate corrective action included testing the hot gas bypass valves to demonstrate the proper performance of their safety function.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the hot gas bypass valve closure is required to ensure the essential chiller can perform its safety function during all design basis accident conditions. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to any external initiating events. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date test procedures were available to demonstrate that equipment performance is adequate to assure nuclear safety [H.2(c)]
Inspection Report# : 2012002 (pdf)
Significance:        Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate the Impact of Fire Damage on the Dry Cooling Tower Fans The team identified a non-cited violation of License Condition 2.C.9 and Appendix R, Section III.G for the failure to adequately evaluate the impact of fire damage on the dry cooling tower fans. Specifically, the failure to adequately evaluate fire damage to the dry cooling tower fans did not ensure one train remained available to achieve and maintain hot shutdown conditions from the alternate shutdown panel. The licensee documented this deficiency in Condition Report 2012-00837.
The failure to adequately evaluate the impact of fire damage on the dry cooling tower fans was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Since this finding involved a control room Page 3 of 13
 
3Q/2012 Inspection Findings - Waterford 3 abandonment issue, a senior reactor analyst performed a Phase 3 significance determination. The senior reactor analyst determined this finding had very low risk significance based upon a bounding analysis (Green). The dominant core damage sequences involved a fire initiating event, failure of both the component cooling water and auxiliary component cooling water systems, as well as an independent failure of the turbine driven auxiliary feedwater pump.
Equipment that helped to mitigate the significance included the unaffected offsite power system, the viable steam generators and the safety related auxiliary feedwater system. Because the original failure to evaluate the impact of fire damage on the dry cooling tower fans had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Significance:      Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate Adequate Cooling Provided to Diesel Generator B within Required Time The team identified a non-cited violation of License Condition 2.C.9 and the fire protection program for the failure to perform a post-fire safe shutdown analysis design calculation. Specifically, the team determined that the licensee had not calculated the time available to establish component cooling water to prevent damaging the emergency diesel generator when providing power to post fire safe shutdown components. The licensee documented this deficiency in Condition Report 2012 00818.
The failure to perform a design calculation evaluating the ability to remove heat based upon emergency diesel generator loading following a control room fire was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Using Appendix F, the team assigned this finding a low degradation rating because the system was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. Specifically, the component cooling water system could accommodate the heat in the jacket water system of a lightly loaded diesel generator. This finding screened as very low safety significance (Green) in the Phase 1 evaluation. Because the original failure to perform a design calculation had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct a Condition Adverse to Quality Associated with the Main Feedwater Isolation Valves The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion XVI because the licensee failed to identify and correct a condition adverse to quality associated with the main feedwater isolation valve. Specifically, the licensee did not identify that varnish deposits were causing the main feedwater isolation valve to fail its inservice testing. As a result, corrective actions that were implemented did not address the adverse condition, leading to a subsequent test failure. The licensee entered this issue into their corrective action program as CRWF3-2011-2005 and CR-WF3-2011-8140. The corrective actions included the replacement of the Page 4 of 13
 
3Q/2012 Inspection Findings - Waterford 3 actuator, a shortening of the replacement frequency of the fourway hydraulic valves to a 36 month interval, and an evaluation of the current methods of gathering and implementing operating experience.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the main feedwater isolation valve is credited for closure during a main feedwater line break. The inspectors performed the initial significance determination using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a.However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment. Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment.
Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves.
Accordingly, the contribution to core damage was much less than E-6. Therefore, the inspectors determined that this finding had very low safety significance (Green). This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not collect and evaluate relevant external operating experience to identify that other sites experienced similar failures of feedwater isolation valves due to varnish deposits on the interior surface [P.2(a)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Tornado Impact on the Ultimate Heat Sink During a Refueling Outage The inspectors identified a non-cited violation of 10CFR50, Appendix B, Criterion III because the licensee did not translate applicable regulatory requirements and the design basis into specifications and instructions.
Specifically, the licensee did not translate the design basis tornado event into a design calculation. This outage-specific calculation was referenced by operations as the basis to ensure that the number of dry cooling tower fans needed for decay heat removal remained available. As a result, additional analysis needed to be performed to verify that the ultimate heat sink would have been able to perform its design function had a design basis tornado occurred during refueling outage RF-17. The licensee entered this issue into their corrective action program as CRWF3-2011-6480. The immediate corrective actions taken to restore compliance included analysis of the condition and actions to ensure that future outage specific calculations include the tornado design basis event.
The performance deficiency is more than minor because it challenges the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating Page 5 of 13
 
3Q/2012 Inspection Findings - Waterford 3 events to prevent undesirable consequences. Since the calculation was used when the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process, and Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Operational Checklists." The issue was determined to have a very low safety significance (Green) because it did not require a quantitative assessment. Through calculation review, the inspectors concluded that this failure resulted in the potential to enter an unanalyzed condition. This finding had a crosscutting aspect in the resources component of the human performance area in that the licensee failed to incorporate accurate design information into instructions [H.2(c)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Work Order Instructions to Install a Swagelok Fitting on a Main Feedwater Isolation Valve Tube Connection The inspectors identified a non-cited violation of Technical Specification 6.8.1.a because the licensee did not follow work order instructions to install a pressure gage in an air line used to measure and maintain pressure for the hydraulic accumulators that close the main feedwater isolation valve. Specifically, the licensee did not follow the instructions to assemble and tighten a Swagelok fitting according to the work order. As a result, the fitting failed, preventing the valve from being able to perform its safety-related function. The licensee entered this issue into their corrective action program as CR-WF3-2010-1166 and CRWF3-2011-7469. The immediate corrective actions included repairing the Swagelok fitting and completing an apparent cause evaluation to determine the nature of the fitting failure and failure to follow procedure.
The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspector performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. However, the main feedwater isolation valves were not included in the pre-solved worksheet and the valves did not appear as components in the Phase 2 significance determination worksheets. The senior reactor analyst performed a Phase 3 significance determination for this issue. The analyst noted that the main feed isolation valves were not a significant contributor to core damage frequency and were not included in the NRCs SPAR model. These valves close to mitigate core overcooling events or to isolate feedwater flow to a ruptured feedwater line inside containment.
Overcooling events do not lead to core damage. A ruptured feedwater line could challenge containment integrity, but without core damage there would be no Page 6 of 13
 
3Q/2012 Inspection Findings - Waterford 3 potential for a large early release. If a valve failed to close on demand, the licensee had other means to isolate feedwater flow to a steam generator or into containment. Operators could secure feedwater pumps, close a block valve, or close the main feedwater flow control valves. Accordingly, the contribution to core damage was much less than E-6. As a result, this finding had a very low safety significance (Green). This finding does not have a crosscutting aspect since it is not indicative of current plant performance.
Inspection Report# : 2011005 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Design Basis Input was Correctly Translated into Design Basis Calculations The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 28, 2011, the licensee failed to assure that design basis information associated with loading the auxiliary component cooling water pumps on the Class 1E Bus was correctly translated in various design basis calculations.
This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2011-06737 and CR-WF3-2011-06808.
The team determined that the failure to verify the adequacy of the design for loading the auxiliary component cooling water pumps on the Class 1E Bus in various design basis calculations was a performance deficiency. 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. Specifically, the inadequate design calculations could have prevented continued operation of the emergency diesel generator under degraded voltage, short circuit, and increased fuel oil consumption conditions. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee revised the associated calculations to include the required 295 brake horsepower value and reanalyzed for verification of operability. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Containment Spray Pump Design Basis Verification Surveillance Test The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Specifically, as of October 4, 2011, the licensee did not have an adequate test procedure to verify containment spray pump design basis accident performance requirements. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06852.
The team determined that the failure to either have a stand-alone design basis accident containment spray pump Page 7 of 13
 
3Q/2012 Inspection Findings - Waterford 3 verification test or to have it adequately incorporated into the in-service testing requirements was a performance deficiency. 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. Specifically, neither the design basis analysis nor related in-service test surveillances, accounted for the inherent uncertainty of the flow element in the overall instrument uncertainty evaluation. In accordance with NRC Inspection Manual Chapter 0609, , "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Extrapolation of Vendor Supplied Pump Net Positive Suction Head Values The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of October 4, 2011, the licensee extrapolated the values for required pump net positive suction head beyond those provided in vendor certified curves without adequate analysis or justification. Consequently, the licensee, per the station-approved net positive suction head analysis, could have operated the safety-related pumps in beyond-analyzed or vendor-approved flow regimes. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06870.
The team determined that the failure to provide adequate justification for extrapolation of net positive suction head values beyond those provided in the certified pump vendor data was a performance deficiency. 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. Specifically, potential pump cavitation at higher than analyzed or vendor-approved operation, could have rendered mitigating equipment (i.e., pumps) to fail. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings,"
the issue was determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed additional analyses to assure that the pumps could safely operate in the required flow regimes. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Adequate Preventive Maintenance Procedures for Aluminum/Copper Electrical Connections to the Ultimate Heat Sink Transformers The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, Activities affecting quality shall be prescribed by documented instructions, Page 8 of 13
 
3Q/2012 Inspection Findings - Waterford 3 procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, as of October 7, 2011, when developing and implementing preventive maintenance procedures and work orders for transformers and electrical connections, the licensee failed to provide specific acceptance criteria and instructions addressing the potential vulnerability of these connections to degradation from galvanic reaction or differential thermal expansion, particularly in a high humidity outdoor environment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06832.
The team determined that the failure to provide suitable acceptance criteria and instructions in preventive maintenance procedures and work orders applicable to the aluminum/copper electrical connections to the transformers was a performance deficiency. 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. Specifically, inadequate preventive maintenance of the aluminum/copper connections could lead to degradation of the electrical connections to the station service transformer and loss of the ultimate heat sink. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis to Support the Adequacy of the Four Inch Bulkhead Drain to Protect the Ultimate Heat Sink During Flood Events The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design bases are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event. Failure of the motor control center as a result of flooding from the probable maximum precipitation event could result in the loss of the associated ultimate heat sink because the motor control center serves both the dry cooling tower and wet cooling tower fan motors. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06701.
The team determined that the failure to establish and maintain an analysis supporting the adequacy of a single four-inch overflow (bulkhead) drain for protecting the ultimate heat sink motor control center from flooding during a design basis probable maximum precipitation event was a performance deficiency. 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. Specifically, the design basis analysis for the four-inch bulkhead drain did not ensure that the motor control center would be adequately protected during a probable maximum precipitation event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a Page 9 of 13
 
3Q/2012 Inspection Findings - Waterford 3 design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed calculations to justify the adequacy of the installed bulkhead drain for the probable maximum precipitation event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Analysis of the Effects of Reverse Rotation of Dry Cooling Tower Fan Motors Resulting from a Tornado Event The team identified a Green violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to October 7, 2011, the licensee failed to analyze the dry cooling tower fan motors for premature trip as a result of reverse rotation caused by a tornado event that could result in the loss of the dry cooling tower heat removal capability. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06850.
The team determined that the failure to establish and maintain an analysis supporting the ability of the dry cooling tower fan motors to operate successfully during and following a design basis tornado event was a performance deficiency. 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. Specifically, the design basis analysis did not ensure that the dry cooling tower fan motors would perform as required under reverse rotation conditions, without premature trip, during a design basis tornado. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee prepared an evaluation of the effect on fan motor starting current and duration for reverse rotation conditions. For reverse rotation conditions that would extend the locked rotor current time by a factor of two, the licensees analysis showed ample margin for the instantaneous trip settings from the magnetic-only breaker and the thermal overload protection, such that premature trip would be precluded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011007 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide an Adequate Basis for Temperature Limits of Auxiliary Component Cooling Water Pump Motor Bearings The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 7, 2011, the licensee did not have an adequate technical basis for increasing the auxiliary component cooling Page 10 of 13
 
3Q/2012 Inspection Findings - Waterford 3 water pump motor bearing temperature alarm setpoints or establishing an upper limit on motor bearing temperature, which directed operators to secure the pump. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06573.
The team determined that the failure to provide an adequate basis for increasing the high bearing temperature alarm setpoints and establishing a high temperature motor trip criterion was a performance deficiency. 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. In accordance with NRC Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed an engineering justification for the bearing temperatures based on industry guidance. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2011007 (pdf)
Significance:      Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage the Risk Involved with a Maintenance Window for the Turbine Driven Essential Feedwater Pump The team identified a Green noncited violation of 10 CFR 50.65(a)(4), which states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, on October 28, 2010 the turbine driven essential feedwater pump was out of service for maintenance for approximately 12 hours. During this time the licensee unknowingly entered the Orange risk window (crossed a risk threshold) due to a faulty assumption in the probabilistic risk assessment model. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2011-06653.
The team determined that the failure to perform adequate risk assessments is a performance deficiency. This finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the issue was identified as requiring a Phase 2 evaluation. A Region IV Senior Reactor Analyst performed a Phase 2 significance determination using NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K:
Delta-CDF = [CCDPActual - CCDPflawed]
* duration /8760 The licensee bounded the duration of the turbine driven essential feedwater pump maintenance at 8 hours in a year.
The flawed ICDP was 3.1E-5, the actual ICDP was 3.1E-5 + 1.9E-5 = 5.0E-5. The difference was 1.9E-5.
Delta-CDF = 1.9E-5
* 12/8760 = 2.6E-8 Therefore, the issue was determined to have very low safety significance (Green). This finding was determined to have a cross-cutting aspect in the area of problem identification and resolution associated with the self and Page 11 of 13
 
3Q/2012 Inspection Findings - Waterford 3 independent assessments component because the licensee performed a probabilistic risk assessment model update in April 2009, which failed to identify the faulty assumption [P.3(a)].
Inspection Report# : 2011007 (pdf)
Barrier Integrity Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inoperable Train of Containment Cooling System The inspectors identified a non-cited violation of Technical Specification Limiting Condition for Operation 3.6.2.2, Containment Cooling System, which requires in Modes 1, 2, 3, and 4 that Two independent trains of containment cooling shall be OPERABLE with one fan cooler to each train. The Technical Specification Action statement requires that With one train of containment cooling inoperable, restore the inoperable train to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable containment cooling train to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours. Specifically, from July 11, 2009, to July 19, 2009, the licensee failed to declare train B of the containment cooling system inoperable, and restore it to operable status within 72 hours or place the unit in hot standby in 6 hours. This finding has been entered into the licensees corrective action program as Condition Reports CR-WF3-2011-08150.
The inspectors determined that the failure to meet Technical Specification Limiting Condition for Operation 3.6.2.2 was a performance deficiency. The finding was more than minor because it adversely affected the structures, systems, and components and barrier performance attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the component cooling water flow for containment cooling system train B decreased below the minimum flow limits of Technical Specification Surveillance Requirement 4.6.2.2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment and heat removal components, and did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. This finding was determined to have a crosscutting aspect in the area of human performance associated with the decision making component because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2011005 (pdf)
Emergency Preparedness Occupational Radiation Safety Page 12 of 13
 
3Q/2012 Inspection Findings - Waterford 3 Public Radiation Safety Significance: SL-IV Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Updated Final Safety Analysis Report The inspectors identified a non-cited violation of 10 CFR 50.71 Maintenance of Records, because the licensee failed to update their updated final safety analysis report with submittals that include a change made to the facility.
Specifically, the licensee built the low level radwaste storage facility in 1995 on the owner controlled area for interim radwaste storage of dry and solidified radioactive waste and failed to update the updated final safety analysis report to include these changes. This issue was entered in the licensees corrective action program as condition report WF3-2011-07711.
This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that stored radwaste materials with a significant radioactive source term has been relocated from the plant radiologically controlled area to the owner controlled area.
In addition, the radwaste management program has been affected because the licensee was not originally licensed to act as a low level waste facility. However, the termination of the Barnwell Low Level Radioactive Waste Management facility has forced the licensee to build such a storage area and make changes to the facility, significantly increasing the onsite storage capacity. The inspectors determined that this finding did not reflect present performance because it is an issue with changes made to the facility more than 15 years previously. Therefore, there was no cross-cutting aspect associated with this finding. This finding is characterized as a Severity Level IV non-cited violation in accordance with NRC Enforcement Policy, Section 6.1, and was treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy Inspection Report# : 2011005 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : November 30, 2012 Page 13 of 13
 
4Q/2012 Inspection Findings - Waterford 3 Waterford 3 4Q/2012 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately access and manage risk before performing maintenance activities associated with non-standard lifts The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4) because the licensee did not assess and manage the overall on line risk involved with maintenance activities that lifted heavy loads over safety related equipment.
Specifically, the licensee did not assess and manage the integrated plant risk prior to performing heavy load lifts in the train B dry cooling tower fan area when installing a temporary work platform to support the steam generator replacement project. As a result, the licensee did not implement additional risk management actions as required by their procedure EN-WM-104, OnLine Risk Assessment. The licensee entered this condition into the corrective action program as CR-WF3-2012-4195 and CR-WF3-2012-4489. The immediate corrective action taken to restore compliance was to re-evaluate and change the integrated risk classification from a normal risk to a high-risk level and implement the required risk management actions.
The failure to adequately assess and manage overall plant risk prior to performing maintenance activities that lifted heavy loads over the train B dry cooling tower fan area was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to identify non-standard lifts over safety related equipment as high risk prevented the licensee from taking additional risk management actions to limit the likelihood of an event that would upset plant stability. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix K Maintenance Risk Assessment and Risk Management Significance Determination Process to determine the significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, a senior reactor analyst determined that the finding was very low safety significance (Green) because the bounding risk deficit was approximately 1E-7/year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision making component of the human performance area because the licensee did not makes safety significant or risk significant decisions using a systematic process to ensure safety was maintained.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Operator Knowledge of Equipment Status The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow Procedure EN-OP-115, Conduct of Operations. Specifically, the licensee failed to ensure that control room operators knew the status of equipment at all times. While interviewing the person responsible for tracking plant deficiencies, the inspectors discovered that the licensee had two separate governing procedures. These two instructions had different definitions for categories of plant deficiencies and different databases for tracking them. The inspectors then interviewed the on-shift operators in the control room and reviewed both databases. The inspectors identified several issues, including lack of knowledge by the control room operators about which procedure to use, failure to track deficiencies in both databases, and inadequate classification of deficiencies.
The inspectors determined that in March 2010, the licensee changed their process for tracking deficiencies to be consistent with their fleet reporting process. However, the licensee did not revise the procedure and did not train all Page 1 of 12
 
4Q/2012 Inspection Findings - Waterford 3 affected personnel on the new process. As a result, control room operators did not have a complete and accurate knowledge of all plant deficiencies. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03732.
The failure to ensure that operators were aware of the status of all plant equipment was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to implement a procedure designed to ensure operators were aware of deficiencies in the instrumentation, controls, and operation of nuclear plant systems. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not cause a reactor trip and did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance area, work practices component, in that the licensee failed to define and effectively communicate expectations regarding procedural compliance, and personnel did not follow procedures.
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Develop Effective Corrective Actions to Preclude Repetition The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, because the licensee failed to determine the cause of a significant condition adverse to quality and take corrective actions to preclude repetition. Specifically, the licensee failed to assure that the cause of the condition was determined and corrective action taken to preclude repetition related to a contractors non-compliance with site procedural requirements. The corrective actions include developing additional training and provisions to provide additional contractor oversight. This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2012-03769 and CR-WF3-2012-03772.
The failure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could lead to more significant consequences; therefore, it is a finding. Specifically, failure to determine the cause of a significant condition adverse to quality and take corrective action to prevent recurrence can result in recurrence of the condition. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance, work practice component, in that the licensee failed to follow guidance in the root cause evaluation procedure when developing appropriate corrective actions to prevent repetition.
Inspection Report# : 2012008 (pdf)
Significance:      Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to establish adequate procedural guidance to control feedwater heater level control valves A self-revealing finding occurred because the licensee did not establish adequate procedural guidance to control feedwater heater level control valves. Specifically, the procedures used to control the settings for the valves did not contain guidance that properly adjusted the proportional gain and air pressure input to ensure the valves open quickly during a transient. As a result, multiple failures in the feedwater heater drain system resulted in a feedwater pump A Page 2 of 12
 
4Q/2012 Inspection Findings - Waterford 3 trip and a subsequent reactor power cutback. The licensee entered this condition into their corrective action program as CR-WF3-2012-1729 for resolution. The corrective actions included a revision of the procedure and loop calibration settings for the feedwater heater level control valves.
The failure to provide adequate guidance that properly adjusted the proportional gain to ensure the valves open as designed is a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, multiple feedwater heater control valve failures resulted in a reactor power cutback that upset plant stability. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date design documentation for loop calibration settings was available to assure nuclear safety [H.2(c)].
Inspection Report# : 2012003 (pdf)
Mitigating Systems Significance:      Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct the cause of repetitive failures associated with train A component cooling water radiation monitor The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, because the licensee did not promptly identify and correct a condition adverse to quality associated with the train A safety-related component cooling water (CCW) radiation monitor (PRMIR7050A). Specifically, the licensee did not identify and correct the cause of repetitive failures of the train A CCW radiation monitor when the monitor experienced erratic radiation spikes and repeat issues with the detector. As a result, the licensee declared the radiation monitor inoperable on several occasions over a span of nine months. The licensee entered this issue into their corrective action program as CR-WF3-2012-4643. The immediate corrective actions taken to restore compliance included the replacement of all susceptible components of the radiation monitor and other associated equipment. Additionally, the licensee adjusted the low-level discriminator voltage and changed the calibration procedure to align testing with vendor recommendations.
The failure to promptly identify and correct the cause of repetitive failures associated with erratic radiation spikes and a repeat issue with the radiation monitor detector was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the erratic radiation spikes and issues with the detector challenged the availability and reliability of the train A CCW radiation monitor used to alert operators of radiation leaks from the reactor coolant system. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, represent a loss of system function, or an actual loss of function of at least a single train for greater than its Tech Spec allowed outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate the problem such that the resolutions address causes and extent of conditions.
Inspection Report# : 2012005 (pdf)
Page 3 of 12
 
4Q/2012 Inspection Findings - Waterford 3 Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical operator actions associated with security and fire events The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a because the licensee did not establish procedural controls to ensure that the assigned minimum staff of licensed operators could perform immediate and time critical operator actions associated with a security or fire event.
Specifically, the licensee did not establish procedural guidance to restrict licensed operators from leaving the PA. As a result, the licensee could not ensure that operators would respond in a timely manner to perform immediate and time critical operator actions required by a fire or security event. The licensee entered this issue into their corrective action program as CR-WF3-2012-3815. The immediate corrective actions taken to restore compliance included the issuing of a standing instruction to instruct the assigned minimum staff of licensed operators to remain in the PA unless officially relieved of their duties.
The failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical steps associated with security and fire events was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee would have been challenged to complete immediate and time critical steps with licensed operators being outside the PA. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it could be risk significant for external events. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.0 E-7/year. The risk important sequences included control room fires that required a control room evacuation. The short duration of the operator being outside the PA helped to reduce the risk significance. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through changes to station processes and procedures.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct conditions adverse to quality related to the header A auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A. Specifically, the licensee did not promptly identify and correct degraded conditions associated with the valves shaft bushings, a pneumatic transducer that controls the valve actuator, and its soft seat. As a result, the licensee declared the valve inoperable on several occasions. The licensee entered this issue into their corrective action program as CR-WF3-2012-03280. The immediate corrective actions taken to restore compliance included the replacement of all the degraded components.
The failure to promptly identify and correct multiple degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded components challenged the closed safety function of the valve and its ability to maintain an adequate water inventory for the wet cooling tower following a loss of coolant accident. The inspector used NRC Inspection Manual 0609, , "Initial Characterization of Findings," to evaluate this issue. The finding required a detailed analysis Page 4 of 12
 
4Q/2012 Inspection Findings - Waterford 3 because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.2E-7 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a torn diaphragm of a safety-related air operated valve associated with the emergency feedwater system The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct a condition adverse to quality associated with the B emergency feedwater backup control valve EFW-223B. Specifically, the licensee did not promptly identify and correct internal leakage from tears in the EFW-223B actuator diaphragm. As a result, these internal tears in the diaphragm caused excessive leakage that affected two nitrogen accumulators used to operate EFW-223B and other safety related valves. The licensee entered this issue into their corrective action program as CR-WF3-2012-0860. The immediate corrective actions taken to restore compliance included the replacement of the diaphragm and to determine the extent of condition for other air-operated valves with the same type, make, and model diaphragm. The planned corrective action included the revision of the air operated valve program post maintenance tests to identify similar problems.
The failure to promptly identify and correct tears in the internal actuator diaphragm of the B emergency feedwater backup control valve EFW-223B was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the internal leakage of EFW-223B affected two safety-related nitrogen accumulators and their ability to provide nitrogen gas to other connected safety related valves following a loss of offsite power event. The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time.
Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding is of very low safety significance (Green) because the bounding change to the core damage frequency is less than 1E-9 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Determine the Operability of the Emergency Diesel Generators The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow the Operability Determination Process. Specifically, the licensee failed to determine the operability of the emergency diesel generators immediately upon discovery without delay and in a controlled manner using the best information available in response to NRC Information Notice 2010-04. The licensee completed an evaluation of the information notice that indicated that Waterford 3 was vulnerable and susceptible to the issue, but the licensee failed to issue a condition report as required by their procedure. The failure to Page 5 of 12
 
4Q/2012 Inspection Findings - Waterford 3 initiate a condition report resulted in the licensees failure to perform an operability determination of the emergency diesel generators as required by, EN-OP-104, Operability Determination Process, Revision 6. In the evaluation, the licensee considered the fact that they had an Action Request in their system to add routine thermography inspections within the voltage regulator cabinets to their preventative maintenance program as being adequate. The action request was not completed when the inspection team reviewed the issue. The inspectors questioned whether there was an operability concern for the emergency diesel generators. The licensee recognized their failure to perform an operability determination. They performed a prompt operability determination based on no observed degradation in performance and declared the emergency diesel generators operable. In addition, they plan to conduct the thermography inspections during the next scheduled emergency diesel generator surveillance. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03761.
The failure to promptly perform an operability determination of the emergency diesel generators in response to NRC Information Notice 2010-04 was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance 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. Specifically, the licensee failed to promptly determine the operability of the diesel generators after obtaining information of a potential condition adverse to quality. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it was not a deficiency affecting the design or qualification of the system, it did not represent a loss of system or function, and it was a Technical Specification system, but did not represent an actual loss of function of a single train for greater than it allowed outage time. Specifically, the licensee performed an operability determination in response to the inspectors questions and determined the emergency diesel generators were operable based on a review of surveillance data and maintenance records. This finding had a cross-cutting aspect in the problem identification and resolution area, operating experience component, in that the licensee failed to systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience.
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Corrective Action Associated with Emergency Feedwater Pump AB The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to take timely corrective action for a condition adverse to quality. Specifically, from May 2011, through August 2012, the licensee failed to restore a degraded condition, which included a corrective action to perform a new design analysis for the emergency feedwater pump AB after the removal of heat trace circuit 1-8C, despite having a reasonable amount of time to complete it. Currently, plant operators are required once per shift to perform temperature verifications of the heat trace to ensure condensation does not form in the steam supply pipe to the turbine driven pump and to maintain emergency feedwater pump AB in an operable but degraded status until the design analysis is complete. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03754.
The team determined that the failure to complete the corrective action of performing a new design analysis to determine if emergency feedwater pump AB required a design modification based on the analysis in a timely manner was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to implement this corrective action could result in reduced reliability of the emergency feedwater pump AB. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it affected the design or qualification of mitigating systems, structures, and components; however, the systems, structures, and components maintained operability. This finding Page 6 of 12
 
4Q/2012 Inspection Findings - Waterford 3 had a cross-cutting aspect in the human performance area, resources component, in that the licensee failed to minimize a long-standing equipment issue adequately to assure nuclear safety.
Inspection Report# : 2012008 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering backlogs low enough to support safety.
Inspection Report# : 2012008 (pdf)
Significance:        Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III because the licensee did not provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis. Specifically, the licensee did not ensure that the design calculation used to determine the required number of wet cooling tower fans needed to operate the plant under normal and design conditions utilized the correct equation. As a result, the incorrect calculation provided reasonable doubt as to the operability of the wet cooling tower fans. The licensee entered this issue into their corrective action program as CR-WF3-2012-1395. The immediate corrective actions taken to restore compliance included a preliminary analysis of the condition and actions to perform a review of the methodology, inputs, and assumptions for the ultimate heat sink thermal performance calculations.
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4Q/2012 Inspection Findings - Waterford 3 The failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the wet cooling tower fans are required to be operable for heat removal following all accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in a loss of operability or functionality of the ultimate heat sink. This finding has a cross-cutting aspect in the decision making component of the human performance area in that the licensee did not conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions [H.1(b)].
Inspection Report# : 2012003 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop Preventive Maintenance Tasks for Critical Limit Switches on Component Cooling Water Inlet Isolation Valves A Green self-revealing, non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a occurred because the licensee did not establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches. Specifically, the licensee had not developed preventive maintenance tasks to lubricate or replace critical limit switches that provide a permissive for the operation of the dry cooling tower fans. As a result, on February 4, 2011, the limit switch on valve CC-135A failed to operate as designed and rendered an entire train of fans inoperable. The licensee entered this condition into their corrective action program as CR-WF3-2011-0679 for resolution. The immediate corrective action included the lubrication of the limit switch and the manual stroking of the valve to obtain free and smooth movement of the degraded equipment. The planned corrective actions included the development of a preventive maintenance task to lubricate and replace the limit switches on a scheduled frequency.
The failure to establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since there was no preventive maintenance task for lubrication and replacement of the equipment, the limit switches can become stuck and render an entire train of dry cooling tower fans inoperable. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. The inspectors also concluded that no cross-cutting aspect is applicable to this finding because the performance deficiency is not reflective of current performance.
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Performance of Safety-Related Valves The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, because the licensee did not identify and perform testing on a safety-related component to demonstrate that it would perform Page 8 of 12
 
4Q/2012 Inspection Findings - Waterford 3 satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform proper testing for the essential chiller hot gas bypass valves RFR-106A, B, and C. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service without performing a test and operability evaluation. The licensee entered this condition into the corrective action program as CR-WF3-2012-0632 and CR-WF3-2012-0659. The immediate corrective action included testing the hot gas bypass valves to demonstrate the proper performance of their safety function.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the hot gas bypass valve closure is required to ensure the essential chiller can perform its safety function during all design basis accident conditions. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to any external initiating events. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date test procedures were available to demonstrate that equipment performance is adequate to assure nuclear safety [H.2(c)]
Inspection Report# : 2012002 (pdf)
Significance:        Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate the Impact of Fire Damage on the Dry Cooling Tower Fans The team identified a non-cited violation of License Condition 2.C.9 and Appendix R, Section III.G for the failure to adequately evaluate the impact of fire damage on the dry cooling tower fans. Specifically, the failure to adequately evaluate fire damage to the dry cooling tower fans did not ensure one train remained available to achieve and maintain hot shutdown conditions from the alternate shutdown panel. The licensee documented this deficiency in Condition Report 2012-00837.
The failure to adequately evaluate the impact of fire damage on the dry cooling tower fans was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Since this finding involved a control room abandonment issue, a senior reactor analyst performed a Phase 3 significance determination. The senior reactor analyst determined this finding had very low risk significance based upon a bounding analysis (Green). The dominant core damage sequences involved a fire initiating event, failure of both the component cooling water and auxiliary component cooling water systems, as well as an independent failure of the turbine driven auxiliary feedwater pump.
Equipment that helped to mitigate the significance included the unaffected offsite power system, the viable steam generators and the safety related auxiliary feedwater system. Because the original failure to evaluate the impact of fire damage on the dry cooling tower fans had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
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4Q/2012 Inspection Findings - Waterford 3 Significance:      Feb 17, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate Adequate Cooling Provided to Diesel Generator B within Required Time The team identified a non-cited violation of License Condition 2.C.9 and the fire protection program for the failure to perform a post-fire safe shutdown analysis design calculation. Specifically, the team determined that the licensee had not calculated the time available to establish component cooling water to prevent damaging the emergency diesel generator when providing power to post fire safe shutdown components. The licensee documented this deficiency in Condition Report 2012 00818.
The failure to perform a design calculation evaluating the ability to remove heat based upon emergency diesel generator loading following a control room fire was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F. The performance deficiency affected the fire protection defense-in depth strategies involving post-fire safe shutdown systems. Using Appendix F, the team assigned this finding a low degradation rating because the system was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. Specifically, the component cooling water system could accommodate the heat in the jacket water system of a lightly loaded diesel generator. This finding screened as very low safety significance (Green) in the Phase 1 evaluation. Because the original failure to perform a design calculation had occurred longer than three years prior to this inspection, this finding did not reflect current licensee performance.
Inspection Report# : 2012007 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:      Oct 23, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area The inspectors reviewed a self-revealing non-cited violation of Technical Specification 6.12.2 which resulted because licensee representatives failed to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area. As immediate corrective action, the workers backed away from the upper guide structure until their dose rate alarms cleared. The upper guide structure lift continued until it was in a safe condition on the stand in the deep end of the refueling pool. Corrective action to prevent recurrence was determined after licensee personnel documented the occurrence in the corrective action program as Condition Report WF3 2012 05571 and performed a root cause evaluation. To address the root cause, the governing procedure will be revised to reflect the establishment of a waterline on the upper guide structure which indicates the highest elevation it can be raised out of the water and maintain an acceptable amount of shielding.
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4Q/2012 Inspection Findings - Waterford 3 The failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area, is a performance deficiency. The performance deficiency is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that it exposed workers to higher than planned dose rates. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure because the inspectors concluded there was no way to construct a scenario in which a minor alteration of circumstances would have resulted in a violation of the Part 20 limits, and (4) the ability to assess dose was not compromised. This finding had a cross-cutting aspect in the human performance area, work control component, in that the licensee did not plan work activities appropriately by incorporating risk insights and job site conditions, such as the effects on job site radiation levels when water shielding was reduced.
Inspection Report# : 2012005 (pdf)
Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 24, 2012 Identified By: NRC Item Type: FIN Finding Waterford Steam Electric Station, Unit 3 - Identification and Resolution of Problems Summary The team reviewed approximately 350 condition reports, work orders, engineering evaluations, root and apparent cause evaluations, and other supporting documentation to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. The team reviewed a sample of system health reports, self-assessments, audits, trending reports and metrics, and various other documents related to the corrective action program.
Based on these reviews, the team concluded that the licensees corrective action program and its other processes to identify and correct nuclear safety problems were adequate to support nuclear safety. However, the team noted at times the licensee staff did not always use the corrective action program for problems that were perceived as minor.
The team also noted several challenges in correcting adverse conditions in a timely manner. Further, the licensee had several long-standing issues, which had been in the corrective action process for over a year without resolution. The licensee appropriately evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. However, there was one example where the licensee failed to enter an information notice into their corrective action program for evaluation of a condition adverse to quality. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee Page 11 of 12
 
4Q/2012 Inspection Findings - Waterford 3 performed effective quality assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. Finally, the team determined that the station continued to maintain a safety-conscious work environment. Employees felt free to raise nuclear safety concerns to the attention of management without fear of retaliation.
Inspection Report# : 2012008 (pdf)
Last modified : February 28, 2013 Page 12 of 12
 
1Q/2013 Inspection Findings - Waterford 3 Waterford 3 1Q/2013 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately access and manage risk before performing maintenance activities associated with non-standard lifts The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4) because the licensee did not assess and manage the overall on line risk involved with maintenance activities that lifted heavy loads over safety related equipment.
Specifically, the licensee did not assess and manage the integrated plant risk prior to performing heavy load lifts in the train B dry cooling tower fan area when installing a temporary work platform to support the steam generator replacement project. As a result, the licensee did not implement additional risk management actions as required by their procedure EN-WM-104, OnLine Risk Assessment. The licensee entered this condition into the corrective action program as CR-WF3-2012-4195 and CR-WF3-2012-4489. The immediate corrective action taken to restore compliance was to re-evaluate and change the integrated risk classification from a normal risk to a high-risk level and implement the required risk management actions.
The failure to adequately assess and manage overall plant risk prior to performing maintenance activities that lifted heavy loads over the train B dry cooling tower fan area was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to identify non-standard lifts over safety related equipment as high risk prevented the licensee from taking additional risk management actions to limit the likelihood of an event that would upset plant stability. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix K Maintenance Risk Assessment and Risk Management Significance Determination Process to determine the significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, a senior reactor analyst determined that the finding was very low safety significance (Green) because the bounding risk deficit was approximately 1E-7/year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision making component of the human performance area because the licensee did not makes safety significant or risk significant decisions using a systematic process to ensure safety was maintained.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Operator Knowledge of Equipment Status The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow Procedure EN-OP-115, Conduct of Operations. Specifically, the licensee failed to ensure that control room operators knew the status of equipment at all times. While interviewing the person responsible for tracking plant deficiencies, the inspectors discovered that the licensee had two separate governing Page 1 of 12
 
1Q/2013 Inspection Findings - Waterford 3 procedures. These two instructions had different definitions for categories of plant deficiencies and different databases for tracking them. The inspectors then interviewed the on-shift operators in the control room and reviewed both databases. The inspectors identified several issues, including lack of knowledge by the control room operators about which procedure to use, failure to track deficiencies in both databases, and inadequate classification of deficiencies.
The inspectors determined that in March 2010, the licensee changed their process for tracking deficiencies to be consistent with their fleet reporting process. However, the licensee did not revise the procedure and did not train all affected personnel on the new process. As a result, control room operators did not have a complete and accurate knowledge of all plant deficiencies. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03732.
The failure to ensure that operators were aware of the status of all plant equipment was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to implement a procedure designed to ensure operators were aware of deficiencies in the instrumentation, controls, and operation of nuclear plant systems. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not cause a reactor trip and did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance area, work practices component, in that the licensee failed to define and effectively communicate expectations regarding procedural compliance, and personnel did not follow procedures.
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Develop Effective Corrective Actions to Preclude Repetition The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, because the licensee failed to determine the cause of a significant condition adverse to quality and take corrective actions to preclude repetition. Specifically, the licensee failed to assure that the cause of the condition was determined and corrective action taken to preclude repetition related to a contractors non-compliance with site procedural requirements. The corrective actions include developing additional training and provisions to provide additional contractor oversight. This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2012-03769 and CR-WF3-2012-03772.
The failure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could lead to more significant consequences; therefore, it is a finding. Specifically, failure to determine the cause of a significant condition adverse to quality and take corrective action to prevent recurrence can result in recurrence of the condition. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance, work practice component, in that the licensee failed to follow guidance in the root cause evaluation procedure when developing appropriate corrective actions to prevent repetition.
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1Q/2013 Inspection Findings - Waterford 3 Inspection Report# : 2012008 (pdf)
Significance:      Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to establish adequate procedural guidance to control feedwater heater level control valves A self-revealing finding occurred because the licensee did not establish adequate procedural guidance to control feedwater heater level control valves. Specifically, the procedures used to control the settings for the valves did not contain guidance that properly adjusted the proportional gain and air pressure input to ensure the valves open quickly during a transient. As a result, multiple failures in the feedwater heater drain system resulted in a feedwater pump A trip and a subsequent reactor power cutback. The licensee entered this condition into their corrective action program as CR-WF3-2012-1729 for resolution. The corrective actions included a revision of the procedure and loop calibration settings for the feedwater heater level control valves.
The failure to provide adequate guidance that properly adjusted the proportional gain to ensure the valves open as designed is a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, multiple feedwater heater control valve failures resulted in a reactor power cutback that upset plant stability. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 -
Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date design documentation for loop calibration settings was available to assure nuclear safety [H.2(c)].
Inspection Report# : 2012003 (pdf)
Mitigating Systems Significance:      Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Control Potential Tornado Borne Missile Hazards The inspectors identified a Green, NCV of Technical Specification 6.8.1.a for failure to follow Procedure OP-901-521, Severe Weather and Flooding, Revision 307. Specifically, on February 25, 2013, the licensee entered the off-normal procedure due to a tornado watch and failed to identify and control potential missile hazards. The licensee has entered this issue into the corrective action program as Condition Report CR-WF3-2013-1590, and is planning corrective actions to determine criteria to identify missile hazards needing controls during severe weather events.
The inspectors concluded that the failure to assess and control potential missile hazards was a performance deficiency.
The inspectors concluded the performance deficiency is more than minor, therefore a finding, because it adversely affected the protection against external factors attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Page 3 of 12
 
1Q/2013 Inspection Findings - Waterford 3 Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 4. A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to include qualitative or quantitative criteria for identification and control of potential missile hazards [H.2(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Determination for Nitrogen Leak in MSIV The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings, for the failure of the licensee to accomplish activities affecting quality in accordance with written procedures. Specifically, operations personnel failed to follow Procedure EN-OP-104, Operability Determinations, and declared main steam isolation valve 1 operable with a through-wall leak on an ASME Class 3 system, despite procedural guidance to the contrary. The licensee has entered this issue into the corrective action program as CR-WF3-2013-1284, and has implemented an ASME Code leak repair as corrective action to restore the degraded condition and reinforced expectations of procedural use and adherence with operations personnel.
The inspectors concluded that the failure of operations personnel to follow procedures was a performance deficiency.
The inspectors determined that the performance deficiency is more than minor, therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 2, A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the component of decision making because the licensee failed to make conservative assumptions when assessing the source of the nitrogen leak and failed to validate underlying assumptions on subsequent operability reviews [H.1(b)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Local Manual Operation Action on Safety-Related Air-Operated Valves The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee failed to identify and perform testing on safety-related components to demonstrate that they would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform testing on several safety-related air-operated valves to demonstrate local manual operation in the event their safety-related nitrogen accumulators were exhausted. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service in accordance with requirements contained in the updated final safety analysis report (UFSAR) and design calculations. The licensee entered this issue into their corrective action program as CR-WF3-2012-6703. The immediate corrective actions taken to restore compliance included developing a test for the local manual operation for some valves and the installation of a backup air supply to recharge the accumulators for other valves.
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1Q/2013 Inspection Findings - Waterford 3 The failure to identify and perform testing to demonstrate that safety-related air-operated valves would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of a system function of safety related equipment. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-8/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included losses of offsite power, which result in an early loss of the instrument air compressors. The fact the accumulators would allow continued air operated valve operation for ten or more hours helped to reduce the risk. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct the cause of repetitive failures associated with train A component cooling water radiation monitor The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, because the licensee did not promptly identify and correct a condition adverse to quality associated with the train A safety-related component cooling water (CCW) radiation monitor (PRMIR7050A). Specifically, the licensee did not identify and correct the cause of repetitive failures of the train A CCW radiation monitor when the monitor experienced erratic radiation spikes and repeat issues with the detector. As a result, the licensee declared the radiation monitor inoperable on several occasions over a span of nine months. The licensee entered this issue into their corrective action program as CR-WF3-2012-4643. The immediate corrective actions taken to restore compliance included the replacement of all susceptible components of the radiation monitor and other associated equipment. Additionally, the licensee adjusted the low-level discriminator voltage and changed the calibration procedure to align testing with vendor recommendations.
The failure to promptly identify and correct the cause of repetitive failures associated with erratic radiation spikes and a repeat issue with the radiation monitor detector was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the erratic radiation spikes and issues with the detector challenged the availability and reliability of the train A CCW radiation monitor used to alert operators of radiation leaks from the reactor coolant system. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, represent a loss of system function, or an actual loss of function of at least a single train for greater than its Tech Spec allowed outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate the problem such that the resolutions address causes and extent of conditions.
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1Q/2013 Inspection Findings - Waterford 3 Inspection Report# : 2012005 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical operator actions associated with security and fire events The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a because the licensee did not establish procedural controls to ensure that the assigned minimum staff of licensed operators could perform immediate and time critical operator actions associated with a security or fire event.
Specifically, the licensee did not establish procedural guidance to restrict licensed operators from leaving the PA. As a result, the licensee could not ensure that operators would respond in a timely manner to perform immediate and time critical operator actions required by a fire or security event. The licensee entered this issue into their corrective action program as CR-WF3-2012-3815. The immediate corrective actions taken to restore compliance included the issuing of a standing instruction to instruct the assigned minimum staff of licensed operators to remain in the PA unless officially relieved of their duties.
The failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical steps associated with security and fire events was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee would have been challenged to complete immediate and time critical steps with licensed operators being outside the PA. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it could be risk significant for external events. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.0 E-7/year. The risk important sequences included control room fires that required a control room evacuation. The short duration of the operator being outside the PA helped to reduce the risk significance. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through changes to station processes and procedures.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct conditions adverse to quality related to the header A auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A. Specifically, the licensee did not promptly identify and correct degraded conditions associated with the valves shaft bushings, a pneumatic transducer that controls the valve actuator, and its soft seat. As a result, the licensee declared the valve inoperable on several occasions. The licensee entered this issue into their corrective action program as CR-WF3-2012-03280. The immediate corrective actions taken to restore compliance included the replacement of all the degraded components.
The failure to promptly identify and correct multiple degraded conditions associated with the auxiliary component Page 6 of 12
 
1Q/2013 Inspection Findings - Waterford 3 cooling water heat exchanger outlet temperature control valve ACC-126A was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded components challenged the closed safety function of the valve and its ability to maintain an adequate water inventory for the wet cooling tower following a loss of coolant accident. The inspector used NRC Inspection Manual 0609, , "Initial Characterization of Findings," to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.2E-7 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary.
Inspection Report# : 2012004 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a torn diaphragm of a safety-related air operated valve associated with the emergency feedwater system The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct a condition adverse to quality associated with the B emergency feedwater backup control valve EFW-223B. Specifically, the licensee did not promptly identify and correct internal leakage from tears in the EFW-223B actuator diaphragm. As a result, these internal tears in the diaphragm caused excessive leakage that affected two nitrogen accumulators used to operate EFW-223B and other safety related valves. The licensee entered this issue into their corrective action program as CR-WF3-2012-0860. The immediate corrective actions taken to restore compliance included the replacement of the diaphragm and to determine the extent of condition for other air-operated valves with the same type, make, and model diaphragm. The planned corrective action included the revision of the air operated valve program post maintenance tests to identify similar problems.
The failure to promptly identify and correct tears in the internal actuator diaphragm of the B emergency feedwater backup control valve EFW-223B was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the internal leakage of EFW-223B affected two safety-related nitrogen accumulators and their ability to provide nitrogen gas to other connected safety related valves following a loss of offsite power event. The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time.
Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding is of very low safety significance (Green) because the bounding change to the core damage frequency is less than 1E-9 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary.
Inspection Report# : 2012004 (pdf)
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1Q/2013 Inspection Findings - Waterford 3 Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Determine the Operability of the Emergency Diesel Generators The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow the Operability Determination Process. Specifically, the licensee failed to determine the operability of the emergency diesel generators immediately upon discovery without delay and in a controlled manner using the best information available in response to NRC Information Notice 2010-04. The licensee completed an evaluation of the information notice that indicated that Waterford 3 was vulnerable and susceptible to the issue, but the licensee failed to issue a condition report as required by their procedure. The failure to initiate a condition report resulted in the licensees failure to perform an operability determination of the emergency diesel generators as required by, EN-OP-104, Operability Determination Process, Revision 6. In the evaluation, the licensee considered the fact that they had an Action Request in their system to add routine thermography inspections within the voltage regulator cabinets to their preventative maintenance program as being adequate. The action request was not completed when the inspection team reviewed the issue. The inspectors questioned whether there was an operability concern for the emergency diesel generators. The licensee recognized their failure to perform an operability determination. They performed a prompt operability determination based on no observed degradation in performance and declared the emergency diesel generators operable. In addition, they plan to conduct the thermography inspections during the next scheduled emergency diesel generator surveillance. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03761.
The failure to promptly perform an operability determination of the emergency diesel generators in response to NRC Information Notice 2010-04 was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance 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. Specifically, the licensee failed to promptly determine the operability of the diesel generators after obtaining information of a potential condition adverse to quality. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it was not a deficiency affecting the design or qualification of the system, it did not represent a loss of system or function, and it was a Technical Specification system, but did not represent an actual loss of function of a single train for greater than it allowed outage time. Specifically, the licensee performed an operability determination in response to the inspectors questions and determined the emergency diesel generators were operable based on a review of surveillance data and maintenance records. This finding had a cross-cutting aspect in the problem identification and resolution area, operating experience component, in that the licensee failed to systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience.
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Corrective Action Associated with Emergency Feedwater Pump AB The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to take timely corrective action for a condition adverse to quality. Specifically, from May 2011, through August 2012, the licensee failed to restore a degraded condition, which included a corrective action to perform a new design analysis for the emergency feedwater pump AB after the removal of heat trace circuit 1-8C, despite having a reasonable amount of time to complete it. Currently, plant operators are required once per shift to perform temperature Page 8 of 12
 
1Q/2013 Inspection Findings - Waterford 3 verifications of the heat trace to ensure condensation does not form in the steam supply pipe to the turbine driven pump and to maintain emergency feedwater pump AB in an operable but degraded status until the design analysis is complete. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03754.
The team determined that the failure to complete the corrective action of performing a new design analysis to determine if emergency feedwater pump AB required a design modification based on the analysis in a timely manner was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to implement this corrective action could result in reduced reliability of the emergency feedwater pump AB. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it affected the design or qualification of mitigating systems, structures, and components; however, the systems, structures, and components maintained operability. This finding had a cross-cutting aspect in the human performance area, resources component, in that the licensee failed to minimize a long-standing equipment issue adequately to assure nuclear safety.
Inspection Report# : 2012008 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
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1Q/2013 Inspection Findings - Waterford 3 Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering backlogs low enough to support safety.
Inspection Report# : 2012008 (pdf)
Significance:      Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III because the licensee did not provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis. Specifically, the licensee did not ensure that the design calculation used to determine the required number of wet cooling tower fans needed to operate the plant under normal and design conditions utilized the correct equation. As a result, the incorrect calculation provided reasonable doubt as to the operability of the wet cooling tower fans. The licensee entered this issue into their corrective action program as CR-WF3-2012-1395. The immediate corrective actions taken to restore compliance included a preliminary analysis of the condition and actions to perform a review of the methodology, inputs, and assumptions for the ultimate heat sink thermal performance calculations.
The failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the wet cooling tower fans are required to be operable for heat removal following all accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in a loss of operability or functionality of the ultimate heat sink. This finding has a cross-cutting aspect in the decision making component of the human performance area in that the licensee did not conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions [H.1(b)].
Inspection Report# : 2012003 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Page 10 of 12
 
1Q/2013 Inspection Findings - Waterford 3 Significance:      Oct 23, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area The inspectors reviewed a self-revealing non-cited violation of Technical Specification 6.12.2 which resulted because licensee representatives failed to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area. As immediate corrective action, the workers backed away from the upper guide structure until their dose rate alarms cleared. The upper guide structure lift continued until it was in a safe condition on the stand in the deep end of the refueling pool. Corrective action to prevent recurrence was determined after licensee personnel documented the occurrence in the corrective action program as Condition Report WF3 2012 05571 and performed a root cause evaluation. To address the root cause, the governing procedure will be revised to reflect the establishment of a waterline on the upper guide structure which indicates the highest elevation it can be raised out of the water and maintain an acceptable amount of shielding.
The failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area, is a performance deficiency. The performance deficiency is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that it exposed workers to higher than planned dose rates. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure because the inspectors concluded there was no way to construct a scenario in which a minor alteration of circumstances would have resulted in a violation of the Part 20 limits, and (4) the ability to assess dose was not compromised. This finding had a cross-cutting aspect in the human performance area, work control component, in that the licensee did not plan work activities appropriately by incorporating risk insights and job site conditions, such as the effects on job site radiation levels when water shielding was reduced.
Inspection Report# : 2012005 (pdf)
Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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1Q/2013 Inspection Findings - Waterford 3 Miscellaneous Significance: N/A Sep 24, 2012 Identified By: NRC Item Type: FIN Finding Waterford Steam Electric Station, Unit 3 - Identification and Resolution of Problems Summary The team reviewed approximately 350 condition reports, work orders, engineering evaluations, root and apparent cause evaluations, and other supporting documentation to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. The team reviewed a sample of system health reports, self-assessments, audits, trending reports and metrics, and various other documents related to the corrective action program.
Based on these reviews, the team concluded that the licensees corrective action program and its other processes to identify and correct nuclear safety problems were adequate to support nuclear safety. However, the team noted at times the licensee staff did not always use the corrective action program for problems that were perceived as minor.
The team also noted several challenges in correcting adverse conditions in a timely manner. Further, the licensee had several long-standing issues, which had been in the corrective action process for over a year without resolution. The licensee appropriately evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. However, there was one example where the licensee failed to enter an information notice into their corrective action program for evaluation of a condition adverse to quality. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. Finally, the team determined that the station continued to maintain a safety-conscious work environment. Employees felt free to raise nuclear safety concerns to the attention of management without fear of retaliation.
Inspection Report# : 2012008 (pdf)
Last modified : June 04, 2013 Page 12 of 12
 
2Q/2013 Inspection Findings - Waterford 3 Waterford 3 2Q/2013 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system following steam generator replacement The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide adequate post modification testing instructions for activities affecting quality to the circumstances that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system.
The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety [P.2.b].
Inspection Report# : 2013003 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately access and manage risk before performing maintenance activities associated with non-standard lifts The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4) because the licensee did not assess and manage the overall on line risk involved with maintenance activities that lifted heavy loads over safety related equipment.
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2Q/2013 Inspection Findings - Waterford 3 Specifically, the licensee did not assess and manage the integrated plant risk prior to performing heavy load lifts in the train B dry cooling tower fan area when installing a temporary work platform to support the steam generator replacement project. As a result, the licensee did not implement additional risk management actions as required by their procedure EN-WM-104, OnLine Risk Assessment. The licensee entered this condition into the corrective action program as CR-WF3-2012-4195 and CR-WF3-2012-4489. The immediate corrective action taken to restore compliance was to re-evaluate and change the integrated risk classification from a normal risk to a high-risk level and implement the required risk management actions.
The failure to adequately assess and manage overall plant risk prior to performing maintenance activities that lifted heavy loads over the train B dry cooling tower fan area was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to identify non-standard lifts over safety related equipment as high risk prevented the licensee from taking additional risk management actions to limit the likelihood of an event that would upset plant stability. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix K Maintenance Risk Assessment and Risk Management Significance Determination Process to determine the significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, a senior reactor analyst determined that the finding was very low safety significance (Green) because the bounding risk deficit was approximately 1E-7/year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision making component of the human performance area because the licensee did not makes safety significant or risk significant decisions using a systematic process to ensure safety was maintained [H.1(a)].
Inspection Report# : 2012004 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Operator Knowledge of Equipment Status The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow Procedure EN-OP-115, Conduct of Operations. Specifically, the licensee failed to ensure that control room operators knew the status of equipment at all times. While interviewing the person responsible for tracking plant deficiencies, the inspectors discovered that the licensee had two separate governing procedures. These two instructions had different definitions for categories of plant deficiencies and different databases for tracking them. The inspectors then interviewed the on-shift operators in the control room and reviewed both databases. The inspectors identified several issues, including lack of knowledge by the control room operators about which procedure to use, failure to track deficiencies in both databases, and inadequate classification of deficiencies.
The inspectors determined that in March 2010, the licensee changed their process for tracking deficiencies to be consistent with their fleet reporting process. However, the licensee did not revise the procedure and did not train all affected personnel on the new process. As a result, control room operators did not have a complete and accurate knowledge of all plant deficiencies. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03732.
The failure to ensure that operators were aware of the status of all plant equipment was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to implement a procedure designed to ensure operators were aware of deficiencies in the instrumentation, controls, and operation of nuclear plant systems. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Page 2 of 14
 
2Q/2013 Inspection Findings - Waterford 3 Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not cause a reactor trip and did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance area, work practices component, in that the licensee failed to define and effectively communicate expectations regarding procedural compliance, and personnel did not follow procedures [H.4(b)].
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Develop Effective Corrective Actions to Preclude Repetition The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, because the licensee failed to determine the cause of a significant condition adverse to quality and take corrective actions to preclude repetition. Specifically, the licensee failed to assure that the cause of the condition was determined and corrective action taken to preclude repetition related to a contractors non-compliance with site procedural requirements. The corrective actions include developing additional training and provisions to provide additional contractor oversight. This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2012-03769 and CR-WF3-2012-03772.
The failure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could lead to more significant consequences; therefore, it is a finding. Specifically, failure to determine the cause of a significant condition adverse to quality and take corrective action to prevent recurrence can result in recurrence of the condition. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the human performance, work practice component, in that the licensee failed to follow guidance in the root cause evaluation procedure when developing appropriate corrective actions to prevent repetition [H.4(b)].
Inspection Report# : 2012008 (pdf)
Mitigating Systems Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the effects of a PMP flooding event on the reactor auxiliary building roof areas that contained safety-rea The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on Page 3 of 14
 
2Q/2013 Inspection Findings - Waterford 3 the reactor auxiliary building (RAB) roof areas. Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012-7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas.
The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significanc [P.1.a]
Inspection Report# : 2013003 (pdf)
Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to submit a licensee event report within 60 days of discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corr The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER.
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2Q/2013 Inspection Findings - Waterford 3 The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency.
However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC's ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Control Potential Tornado Borne Missile Hazards The inspectors identified a Green, NCV of Technical Specification 6.8.1.a for failure to follow Procedure OP-901-521, Severe Weather and Flooding, Revision 307. Specifically, on February 25, 2013, the licensee entered the off-normal procedure due to a tornado watch and failed to identify and control potential missile hazards. The licensee has entered this issue into the corrective action program as Condition Report CR-WF3-2013-1590, and is planning corrective actions to determine criteria to identify missile hazards needing controls during severe weather events.
The inspectors concluded that the failure to assess and control potential missile hazards was a performance deficiency.
The inspectors concluded the performance deficiency is more than minor, therefore a finding, because it adversely affected the protection against external factors attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 4. A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to include qualitative or quantitative criteria for identification and control of potential missile hazards [H.2(c)].
Inspection Report# : 2013002 (pdf)
Significance:      Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Determination for Nitrogen Leak in MSIV The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings, for the failure of the licensee to accomplish activities affecting quality in accordance with written procedures. Specifically, operations personnel failed to follow Procedure EN-OP-104, Operability Determinations, and declared main steam isolation valve 1 operable with a through-wall leak on an ASME Class 3 system, despite procedural guidance to the contrary. The licensee has entered this issue into the corrective action program as CR-WF3-2013-1284, and has implemented an ASME Code leak repair as corrective action to restore the degraded condition and reinforced expectations of procedural use and adherence with operations personnel.
The inspectors concluded that the failure of operations personnel to follow procedures was a performance deficiency.
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2Q/2013 Inspection Findings - Waterford 3 The inspectors determined that the performance deficiency is more than minor, therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 2, A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the component of decision making because the licensee failed to make conservative assumptions when assessing the source of the nitrogen leak and failed to validate underlying assumptions on subsequent operability reviews [H.1(b)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Local Manual Operation Action on Safety-Related Air-Operated Valves The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee failed to identify and perform testing on safety-related components to demonstrate that they would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform testing on several safety-related air-operated valves to demonstrate local manual operation in the event their safety-related nitrogen accumulators were exhausted. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service in accordance with requirements contained in the updated final safety analysis report (UFSAR) and design calculations. The licensee entered this issue into their corrective action program as CR-WF3-2012-6703. The immediate corrective actions taken to restore compliance included developing a test for the local manual operation for some valves and the installation of a backup air supply to recharge the accumulators for other valves.
The failure to identify and perform testing to demonstrate that safety-related air-operated valves would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of a system function of safety related equipment. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-8/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included losses of offsite power, which result in an early loss of the instrument air compressors. The fact the accumulators would allow continued air operated valve operation for ten or more hours helped to reduce the risk. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Page 6 of 14
 
2Q/2013 Inspection Findings - Waterford 3 Item Type: NCV NonCited Violation Failure to promptly identify and correct the cause of repetitive failures associated with train A component cooling water radiation monitor The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, because the licensee did not promptly identify and correct a condition adverse to quality associated with the train A safety-related component cooling water (CCW) radiation monitor (PRMIR7050A). Specifically, the licensee did not identify and correct the cause of repetitive failures of the train A CCW radiation monitor when the monitor experienced erratic radiation spikes and repeat issues with the detector. As a result, the licensee declared the radiation monitor inoperable on several occasions over a span of nine months. The licensee entered this issue into their corrective action program as CR-WF3-2012-4643. The immediate corrective actions taken to restore compliance included the replacement of all susceptible components of the radiation monitor and other associated equipment. Additionally, the licensee adjusted the low-level discriminator voltage and changed the calibration procedure to align testing with vendor recommendations.
The failure to promptly identify and correct the cause of repetitive failures associated with erratic radiation spikes and a repeat issue with the radiation monitor detector was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the erratic radiation spikes and issues with the detector challenged the availability and reliability of the train A CCW radiation monitor used to alert operators of radiation leaks from the reactor coolant system. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, represent a loss of system function, or an actual loss of function of at least a single train for greater than its Tech Spec allowed outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate the problem such that the resolutions address causes and extent of conditions [P.1(c)].
Inspection Report# : 2012005 (pdf)
Significance:        Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical operator actions associated with security and fire events The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a because the licensee did not establish procedural controls to ensure that the assigned minimum staff of licensed operators could perform immediate and time critical operator actions associated with a security or fire event.
Specifically, the licensee did not establish procedural guidance to restrict licensed operators from leaving the PA. As a result, the licensee could not ensure that operators would respond in a timely manner to perform immediate and time critical operator actions required by a fire or security event. The licensee entered this issue into their corrective action program as CR-WF3-2012-3815. The immediate corrective actions taken to restore compliance included the issuing of a standing instruction to instruct the assigned minimum staff of licensed operators to remain in the PA unless officially relieved of their duties.
The failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical steps associated with security and fire events was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and Page 7 of 14
 
2Q/2013 Inspection Findings - Waterford 3 affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee would have been challenged to complete immediate and time critical steps with licensed operators being outside the PA. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it could be risk significant for external events. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.0 E-7/year. The risk important sequences included control room fires that required a control room evacuation. The short duration of the operator being outside the PA helped to reduce the risk significance. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through changes to station processes and procedures [P.2(b)].
Inspection Report# : 2012004 (pdf)
Significance:      Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct conditions adverse to quality related to the header A auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A. Specifically, the licensee did not promptly identify and correct degraded conditions associated with the valves shaft bushings, a pneumatic transducer that controls the valve actuator, and its soft seat. As a result, the licensee declared the valve inoperable on several occasions. The licensee entered this issue into their corrective action program as CR-WF3-2012-03280. The immediate corrective actions taken to restore compliance included the replacement of all the degraded components.
The failure to promptly identify and correct multiple degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded components challenged the closed safety function of the valve and its ability to maintain an adequate water inventory for the wet cooling tower following a loss of coolant accident. The inspector used NRC Inspection Manual 0609, , "Initial Characterization of Findings," to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.2E-7 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2012004 (pdf)
Significance:      Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a torn diaphragm of a safety-related air operated valve associated with the Page 8 of 14
 
2Q/2013 Inspection Findings - Waterford 3 emergency feedwater system The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct a condition adverse to quality associated with the B emergency feedwater backup control valve EFW-223B. Specifically, the licensee did not promptly identify and correct internal leakage from tears in the EFW-223B actuator diaphragm. As a result, these internal tears in the diaphragm caused excessive leakage that affected two nitrogen accumulators used to operate EFW-223B and other safety related valves. The licensee entered this issue into their corrective action program as CR-WF3-2012-0860. The immediate corrective actions taken to restore compliance included the replacement of the diaphragm and to determine the extent of condition for other air-operated valves with the same type, make, and model diaphragm. The planned corrective action included the revision of the air operated valve program post maintenance tests to identify similar problems.
The failure to promptly identify and correct tears in the internal actuator diaphragm of the B emergency feedwater backup control valve EFW-223B was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the internal leakage of EFW-223B affected two safety-related nitrogen accumulators and their ability to provide nitrogen gas to other connected safety related valves following a loss of offsite power event. The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time.
Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding is of very low safety significance (Green) because the bounding change to the core damage frequency is less than 1E-9 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2012004 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Determine the Operability of the Emergency Diesel Generators The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow the Operability Determination Process. Specifically, the licensee failed to determine the operability of the emergency diesel generators immediately upon discovery without delay and in a controlled manner using the best information available in response to NRC Information Notice 2010-04. The licensee completed an evaluation of the information notice that indicated that Waterford 3 was vulnerable and susceptible to the issue, but the licensee failed to issue a condition report as required by their procedure. The failure to initiate a condition report resulted in the licensees failure to perform an operability determination of the emergency diesel generators as required by, EN-OP-104, Operability Determination Process, Revision 6. In the evaluation, the licensee considered the fact that they had an Action Request in their system to add routine thermography inspections within the voltage regulator cabinets to their preventative maintenance program as being adequate. The action request was not completed when the inspection team reviewed the issue. The inspectors questioned whether there was an operability concern for the emergency diesel generators. The licensee recognized their failure to perform an operability determination. They performed a prompt operability determination based on no observed degradation in performance and declared the emergency diesel generators operable. In addition, they plan to conduct the thermography inspections during the next scheduled emergency diesel generator surveillance. This finding was Page 9 of 14
 
2Q/2013 Inspection Findings - Waterford 3 entered into the licensees corrective action program as Condition Report CR-WF3-2012-03761.
The failure to promptly perform an operability determination of the emergency diesel generators in response to NRC Information Notice 2010-04 was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance 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. Specifically, the licensee failed to promptly determine the operability of the diesel generators after obtaining information of a potential condition adverse to quality. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it was not a deficiency affecting the design or qualification of the system, it did not represent a loss of system or function, and it was a Technical Specification system, but did not represent an actual loss of function of a single train for greater than it allowed outage time. Specifically, the licensee performed an operability determination in response to the inspectors questions and determined the emergency diesel generators were operable based on a review of surveillance data and maintenance records. This finding had a cross-cutting aspect in the problem identification and resolution area, operating experience component, in that the licensee failed to systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience [P.2(a)].
Inspection Report# : 2012008 (pdf)
Significance:      Sep 24, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Corrective Action Associated with Emergency Feedwater Pump AB The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to take timely corrective action for a condition adverse to quality. Specifically, from May 2011, through August 2012, the licensee failed to restore a degraded condition, which included a corrective action to perform a new design analysis for the emergency feedwater pump AB after the removal of heat trace circuit 1-8C, despite having a reasonable amount of time to complete it. Currently, plant operators are required once per shift to perform temperature verifications of the heat trace to ensure condensation does not form in the steam supply pipe to the turbine driven pump and to maintain emergency feedwater pump AB in an operable but degraded status until the design analysis is complete. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2012-03754.
The team determined that the failure to complete the corrective action of performing a new design analysis to determine if emergency feedwater pump AB required a design modification based on the analysis in a timely manner was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to implement this corrective action could result in reduced reliability of the emergency feedwater pump AB. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it affected the design or qualification of mitigating systems, structures, and components; however, the systems, structures, and components maintained operability. This finding had a cross-cutting aspect in the human performance area, resources component, in that the licensee failed to minimize a long-standing equipment issue adequately to assure nuclear safety [H.2(a)].
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2Q/2013 Inspection Findings - Waterford 3 Inspection Report# : 2012008 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering backlogs low enough to support safety [H.2(a)].
Inspection Report# : 2012008 (pdf)
Barrier Integrity Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis Report The inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the Page 11 of 14
 
2Q/2013 Inspection Findings - Waterford 3 licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report, Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193.
The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the Updated Final Safety Analysis Report in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This non-cited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Action 4 of TS 3.3.1 during shutdown with the protective system trip breakers in the open position for Modes 4 and 5 The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that involved reactivity changes use to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not account for temperature increases with a positive MTC within the shutdown margin calculation. This had the potential to affect the available shutdown margin. The licensee entered this issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3-2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service.
The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions Page 12 of 14
 
2Q/2013 Inspection Findings - Waterford 3
[H.1.a].
Inspection Report# : 2013003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Oct 23, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area The inspectors reviewed a self-revealing non-cited violation of Technical Specification 6.12.2 which resulted because licensee representatives failed to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area. As immediate corrective action, the workers backed away from the upper guide structure until their dose rate alarms cleared. The upper guide structure lift continued until it was in a safe condition on the stand in the deep end of the refueling pool. Corrective action to prevent recurrence was determined after licensee personnel documented the occurrence in the corrective action program as Condition Report WF3 2012 05571 and performed a root cause evaluation. To address the root cause, the governing procedure will be revised to reflect the establishment of a waterline on the upper guide structure which indicates the highest elevation it can be raised out of the water and maintain an acceptable amount of shielding.
The failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area, is a performance deficiency. The performance deficiency is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that it exposed workers to higher than planned dose rates. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure because the inspectors concluded there was no way to construct a scenario in which a minor alteration of circumstances would have resulted in a violation of the Part 20 limits, and (4) the ability to assess dose was not compromised. This finding had a cross-cutting aspect in the human performance area, work control component, in that the licensee did not plan work activities appropriately by incorporating risk insights and job site conditions, such as the effects on job site radiation levels when water shielding was reduced [H.3(a)].
Inspection Report# : 2012005 (pdf)
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2Q/2013 Inspection Findings - Waterford 3 Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 24, 2012 Identified By: NRC Item Type: FIN Finding Waterford Steam Electric Station, Unit 3 - Identification and Resolution of Problems Summary The team reviewed approximately 350 condition reports, work orders, engineering evaluations, root and apparent cause evaluations, and other supporting documentation to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. The team reviewed a sample of system health reports, self-assessments, audits, trending reports and metrics, and various other documents related to the corrective action program.
Based on these reviews, the team concluded that the licensees corrective action program and its other processes to identify and correct nuclear safety problems were adequate to support nuclear safety. However, the team noted at times the licensee staff did not always use the corrective action program for problems that were perceived as minor.
The team also noted several challenges in correcting adverse conditions in a timely manner. Further, the licensee had several long-standing issues, which had been in the corrective action process for over a year without resolution. The licensee appropriately evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. However, there was one example where the licensee failed to enter an information notice into their corrective action program for evaluation of a condition adverse to quality. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. Finally, the team determined that the station continued to maintain a safety-conscious work environment. Employees felt free to raise nuclear safety concerns to the attention of management without fear of retaliation.
Inspection Report# : 2012008 (pdf)
Last modified : September 03, 2013 Page 14 of 14
 
3Q/2013 Inspection Findings - Waterford 3 Waterford 3 3Q/2013 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system following steam generator replacement The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide adequate post modification testing instructions for activities affecting quality to the circumstances that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system.
The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety [P.2.b].
Inspection Report# : 2013003 (pdf)
Mitigating Systems Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Page 1 of 11
 
3Q/2013 Inspection Findings - Waterford 3 Failure to implement fire protection program procedure requirements when securing from a fire watch The inspectors identified a non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9, because the licensee did not implement fire protection procedure FP-001-014, Duties of a Fire Watch. Specifically, the licensees fire watch personnel did not implement section 6.5 of FP-001-014 to remove firefighting equipment from work areas when securing from a fire watch. As a result, multiple undercharged fire extinguishers were left in the fire area. The licensee entered this condition into their corrective action program as CR-WF3-2013-03398 and CR-WF3-2013-03523 for resolution. The immediate corrective actions taken to restore compliance included the removal of all undercharged fire watch extinguishers from deactivated posts and returning them to their proper storage location.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to remove undercharged fire extinguishers from work areas that contained safe shutdown equipment could hinder responses to fires in the fire area. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix F Fire Protection Significance Determination Process to determine the significance of the finding. The inspectors determined that the finding had a low degradation rating because it reflected a fire protection program element whose performance and reliability would be minimally impacted.
Specifically, in all cases identified, there were permanent fully charged portable fire extinguishers of the proper type nearby. Therefore, the finding was of very low safety significance (Green). The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to develop preventive maintenance tasks to inspect essential chiller piping for corrosion A self-revealing non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a, occurred because Waterford did not establish preventive maintenance schedules to inspect the safety-related essential chiller water system piping for corrosion. Specifically, the licensee did not establish preventive maintenance task that would inspect the piping associated with the essential chillers. As a result, essential chiller B experienced a through wall piping leak that rendered the chiller inoperable. The licensee entered this condition into their corrective action program as CR-WF3-2013-2876. The immediate corrective actions taken to restore compliance included the replacement of the corroded length of piping alone with an extent of condition walkdown and inspection of the other two remaining chillers.
The failure to establish preventive maintenance schedules to inspect safety-related essential chiller water system piping for corrosion was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the chilled water system, which respond to initiating events to prevent undesirable consequences. Specifically, the essential chilled water system removes heat loads from selected safety-related air handling units during a design basis accident or transient. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, Significance Determination Process (SDP) for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not affect the design or qualification of a mitigating structure, system, or component (SSC), represent a loss of system function, or an actual loss of function of Page 2 of 11
 
3Q/2013 Inspection Findings - Waterford 3 at least a single train for greater than its Technical Specifications allowed outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety [P.2.b].
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to accomplish activities affecting quality on a degraded safety-related solenoid valve in accordance with procedure requirements The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not accomplish activities affecting quality on a degraded safety-related train B component cooling water (CCW) bypass valve (CC-134B) in accordance with maintenance procedure EN-MA-101, Fundamentals of Maintenance. Specifically, the licensee did not control and perform test on a leaking safety-related bypass valve (CC-134B) after maintenance personnel removed the degraded equipment from service as required by Section 5.10 of EN-MA-101. As a result, the licensee could not characterize and determine the cause of the leakage for the safety-related valve. The licensee entered this condition into their corrective action program as CR-WF3-2012-05991, CR-WF3-2012-06288, and CR-WF3-2013-04047. The immediate corrective actions taken to restore compliance included the installation of a new valve and debriefing personnel about controlling equipment removed from service when combining preventative and corrective maintenance tasks in one work order.
The failure to control failed equipment removed from the plant to determine the cause in accordance with maintenance procedure requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition challenged the safety function of the valve (CC-134B) to limit the loss of CCW through damaged portions of the dry cooling tower fans following a tornado-generated missile strike. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it was potentially risk significant for an external event (tornado). Therefore, the senior reactor analyst (SRA) performed a bounding detailed risk evaluation. The SRA determined that the finding was of very low safety significance (Green).
The bounding change to the core damage frequency was less than 3E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of offsite power, failure of the dry cooling tower pressure boundary, failure to isolate the damaged dry cooling tower, and failure to recover instrument air. The redundant train A component cooling water system combined with the tornado frequency helped to reduce the risk exposure. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work control component of the human performance area in that the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope or activity on plant and human performance.
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct a loose flapper set screw associated with an emergency feed water flow control valve Page 3 of 11
 
3Q/2013 Inspection Findings - Waterford 3 A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI occurred because the licensee did not promptly identify and correct a condition adverse to quality related to an emergency feedwater (EFW) backup flow control valve (EFW-223A). Specifically, the licensee did not promptly identify and correct a degraded condition associated with a loose flapper set screw internal to the valve. As a result, the valve failed to perform its safety-related close function when called upon after a plant trip. The licensee entered this condition into their corrective action program as CR-WF3-2013-00451. The immediate corrective actions taken to restore compliance included tightening the loose flapper set screw and recalibrating the valve such that it would close on demand.
The failure to promptly identify and correct a loose flapper set screw associated with EFW-223A was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition challenged the safety function of the valve to prevent feeding a faulted steam generator, and to limit the EFW flow rate injected into the steam generator to minimize the effects of overcooling the reactor coolant system. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, Significance Determination Process (SDP) for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, did not represent a loss of system or function, did not represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) allowed outage time, and did not represent an actual loss of function of one or more non TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not effectively communicate expectations regarding procedural compliance.
Inspection Report# : 2013004 (pdf)
Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the effects of a PMP flooding event on the reactor auxiliary building roof areas that contained safety-rea The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on the reactor auxiliary building (RAB) roof areas. Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012-7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas.
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3Q/2013 Inspection Findings - Waterford 3 The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significanc [P.1.a]
Inspection Report# : 2013003 (pdf)
Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to submit a licensee event report within 60 days of discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corr The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER.
The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency.
However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC's ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:        Mar 31, 2013 Page 5 of 11
 
3Q/2013 Inspection Findings - Waterford 3 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Control Potential Tornado Borne Missile Hazards The inspectors identified a Green, NCV of Technical Specification 6.8.1.a for failure to follow Procedure OP-901-521, Severe Weather and Flooding, Revision 307. Specifically, on February 25, 2013, the licensee entered the off-normal procedure due to a tornado watch and failed to identify and control potential missile hazards. The licensee has entered this issue into the corrective action program as Condition Report CR-WF3-2013-1590, and is planning corrective actions to determine criteria to identify missile hazards needing controls during severe weather events.
The inspectors concluded that the failure to assess and control potential missile hazards was a performance deficiency.
The inspectors concluded the performance deficiency is more than minor, therefore a finding, because it adversely affected the protection against external factors attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 4. A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to include qualitative or quantitative criteria for identification and control of potential missile hazards [H.2(c)].
Inspection Report# : 2013002 (pdf)
Significance:      Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Determination for Nitrogen Leak in MSIV The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings, for the failure of the licensee to accomplish activities affecting quality in accordance with written procedures. Specifically, operations personnel failed to follow Procedure EN-OP-104, Operability Determinations, and declared main steam isolation valve 1 operable with a through-wall leak on an ASME Class 3 system, despite procedural guidance to the contrary. The licensee has entered this issue into the corrective action program as CR-WF3-2013-1284, and has implemented an ASME Code leak repair as corrective action to restore the degraded condition and reinforced expectations of procedural use and adherence with operations personnel.
The inspectors concluded that the failure of operations personnel to follow procedures was a performance deficiency.
The inspectors determined that the performance deficiency is more than minor, therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 2, A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the component of decision making because the licensee failed to make conservative assumptions when assessing the source of the nitrogen leak and failed to validate underlying assumptions on subsequent operability reviews [H.1(b)].
Inspection Report# : 2013002 (pdf)
Significance:      Mar 31, 2013 Page 6 of 11
 
3Q/2013 Inspection Findings - Waterford 3 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Local Manual Operation Action on Safety-Related Air-Operated Valves The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee failed to identify and perform testing on safety-related components to demonstrate that they would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform testing on several safety-related air-operated valves to demonstrate local manual operation in the event their safety-related nitrogen accumulators were exhausted. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service in accordance with requirements contained in the updated final safety analysis report (UFSAR) and design calculations. The licensee entered this issue into their corrective action program as CR-WF3-2012-6703. The immediate corrective actions taken to restore compliance included developing a test for the local manual operation for some valves and the installation of a backup air supply to recharge the accumulators for other valves.
The failure to identify and perform testing to demonstrate that safety-related air-operated valves would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of a system function of safety related equipment. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-8/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included losses of offsite power, which result in an early loss of the instrument air compressors. The fact the accumulators would allow continued air operated valve operation for ten or more hours helped to reduce the risk. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify and correct the cause of repetitive failures associated with train A component cooling water radiation monitor The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, because the licensee did not promptly identify and correct a condition adverse to quality associated with the train A safety-related component cooling water (CCW) radiation monitor (PRMIR7050A). Specifically, the licensee did not identify and correct the cause of repetitive failures of the train A CCW radiation monitor when the monitor experienced erratic radiation spikes and repeat issues with the detector. As a result, the licensee declared the radiation monitor inoperable on several occasions over a span of nine months. The licensee entered this issue into their corrective action program as CR-WF3-2012-4643. The immediate corrective actions taken to restore compliance included the replacement of all susceptible components of the radiation monitor and other associated equipment. Additionally, the licensee adjusted the low-level discriminator voltage and changed the calibration procedure to align testing with vendor recommendations.
The failure to promptly identify and correct the cause of repetitive failures associated with erratic radiation spikes and Page 7 of 11
 
3Q/2013 Inspection Findings - Waterford 3 a repeat issue with the radiation monitor detector was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the erratic radiation spikes and issues with the detector challenged the availability and reliability of the train A CCW radiation monitor used to alert operators of radiation leaks from the reactor coolant system. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, represent a loss of system function, or an actual loss of function of at least a single train for greater than its Tech Spec allowed outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate the problem such that the resolutions address causes and extent of conditions [P.1(c)].
Inspection Report# : 2012005 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering Page 8 of 11
 
3Q/2013 Inspection Findings - Waterford 3 backlogs low enough to support safety [H.2(a)].
Inspection Report# : 2012008 (pdf)
Barrier Integrity Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis Report The inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report, Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193.
The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the Updated Final Safety Analysis Report in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This non-cited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Action 4 of TS 3.3.1 during shutdown with the protective system trip breakers in the open position for Modes 4 and 5 The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that involved reactivity changes use to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not account for temperature increases with a positive MTC within the shutdown margin calculation. This had the potential to affect the available shutdown margin. The licensee entered this issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3-2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service.
The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was Page 9 of 11
 
3Q/2013 Inspection Findings - Waterford 3 more than minor because it was associated with the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions
[H.1.a].
Inspection Report# : 2013003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Oct 23, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area The inspectors reviewed a self-revealing non-cited violation of Technical Specification 6.12.2 which resulted because licensee representatives failed to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area. As immediate corrective action, the workers backed away from the upper guide structure until their dose rate alarms cleared. The upper guide structure lift continued until it was in a safe condition on the stand in the deep end of the refueling pool. Corrective action to prevent recurrence was determined after licensee personnel documented the occurrence in the corrective action program as Condition Report WF3 2012 05571 and performed a root cause evaluation. To address the root cause, the governing procedure will be revised to reflect the establishment of a waterline on the upper guide structure which indicates the highest elevation it can be raised out of the water and maintain an acceptable amount of shielding.
The failure to implement control measures to ensure that activated materials were not raised above or brought near the surface of the refueling pool, causing a locked high radiation area, is a performance deficiency. The performance deficiency is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that it exposed workers to higher than planned dose rates. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined Page 10 of 11
 
3Q/2013 Inspection Findings - Waterford 3 the finding had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure because the inspectors concluded there was no way to construct a scenario in which a minor alteration of circumstances would have resulted in a violation of the Part 20 limits, and (4) the ability to assess dose was not compromised. This finding had a cross-cutting aspect in the human performance area, work control component, in that the licensee did not plan work activities appropriately by incorporating risk insights and job site conditions, such as the effects on job site radiation levels when water shielding was reduced [H.3(a)].
Inspection Report# : 2012005 (pdf)
Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : December 03, 2013 Page 11 of 11
 
4Q/2013 Inspection Findings - Waterford 3 Waterford 3 4Q/2013 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system following steam generator replacement The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide adequate post modification testing instructions for activities affecting quality to the circumstances that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system.
The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety [P.2.b].
Inspection Report# : 2013003 (pdf)
Mitigating Systems Significance:        Dec 20, 2013 Identified By: NRC Item Type: AV Apparent Violation Page 1 of 10
 
4Q/2013 Inspection Findings - Waterford 3 Failure to establish an adequate test program to demonstrate that the train B EDG exhaust fan would perform satisfactorily in service Inspection Report# : 2013008 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Fire Protection Program Procedure Requirements When Securing from a Fire Watch The inspectors identified a non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9, because the licensee did not implement fire protection procedure FP-001-014, Duties of a Fire Watch. Specifically, the licensees fire watch personnel did not implement Section 6.5 of FP-001-014 to remove firefighting equipment from work areas when securing from a fire watch. As a result, multiple undercharged fire extinguishers were left in a fire area. The inspectors determined that this would affect safety-related equipment because it would delay the response to fires in the fire areas. The licensee entered this condition into their corrective action program as CR-WF3-2013-03398 and CR WF3-2013-03523 for resolution. The immediate corrective actions taken to restore compliance included the removal of all undercharged fire extinguishers from deactivated posts and returning them to their proper storage location.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to remove undercharged fire extinguishers from work areas that contained safe shutdown equipment could hinder responses to fires in the fire area. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, to determine the significance of the finding. The inspectors determined that the finding had a low degradation rating because it reflected a fire protection program element whose performance and reliability would be minimally impacted.
Specifically, in all cases identified, there were permanent fully charged portable fire extinguishers of the proper type nearby. Therefore, the finding was of very low safety significance (Green). The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to accomplish activities affecting quality on a degraded safety-related solenoid valve in accordance with procedure requirements The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not accomplish activities affecting quality on a degraded safety-related train B component cooling water (CCW) bypass valve (CC-134B) in accordance with maintenance procedure EN-MA-101, Fundamentals of Maintenance. Specifically, the licensee did not control and perform testing on a leaking solenoid valve related to the operation of a safety-related bypass valve (CC-134B) after maintenance personnel removed the degraded equipment from service as required by Section 5.10 of EN-MA-101. As a result, the licensee could not characterize and determine the cause of the leakage for the safety-related valve. The inspectors determined that this Page 2 of 10
 
4Q/2013 Inspection Findings - Waterford 3 would challenge the safety function of the valve to provide CCW to the ultimate heat sink following a tornado event.
The licensee entered this condition into their corrective action program as CR-WF3-2012-05991, CR-WF3-2012-06288, and CR WF3-2013-04047. The immediate corrective actions taken to restore compliance included the installation of a new valve and debriefing personnel about controlling equipment removed from service when combining preventative and corrective maintenance tasks in one work order.
The failure to control failed equipment removed from the plant to determine the cause in accordance with maintenance procedure requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition challenged the safety function of the valve (CC-134B) to limit the loss of CCW through damaged portions of the dry cooling tower fans following a tornado-generated missile strike. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it was potentially risk significant for an external event (tornado). Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The senior reactor analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 3E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of offsite power, failure of the dry cooling tower pressure boundary, failure to isolate the damaged dry cooling tower, and failure to recover instrument air. The redundant train A component cooling water system combined with the tornado frequency helped to reduce the risk exposure. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work control component of the human performance area in that the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope or activity on plant and human performance [H.3(b)].
Inspection Report# : 2013004 (pdf)
Significance:        Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the effects of a PMP flooding event on the reactor auxiliary building roof areas that contained safety-rea The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on the reactor auxiliary building (RAB) roof areas.
Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012-7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas.
The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that Page 3 of 10
 
4Q/2013 Inspection Findings - Waterford 3 respond to initiating events to prevent undesirable consequences. Specifically, the safety-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1.a].
Inspection Report# : 2013003 (pdf)
Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to submit a licensee event report within 60 days of discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corr The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER.
The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency.
However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC's ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Control Potential Tornado Borne Missile Hazards The inspectors identified a Green, NCV of Technical Specification 6.8.1.a for failure to follow Procedure OP-901-521, Severe Weather and Flooding, Revision 307. Specifically, on February 25, 2013, the licensee entered the off-normal procedure due to a tornado watch and failed to identify and control potential missile hazards. The licensee has entered this issue into the corrective action program as Condition Report CR-WF3-2013-1590, and is planning corrective actions to determine criteria to identify missile hazards needing controls during severe weather events.
The inspectors concluded that the failure to assess and control potential missile hazards was a performance deficiency.
Page 4 of 10
 
4Q/2013 Inspection Findings - Waterford 3 The inspectors concluded the performance deficiency is more than minor, therefore a finding, because it adversely affected the protection against external factors attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 4. A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to include qualitative or quantitative criteria for identification and control of potential missile hazards [H.2(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Operability Determination for Nitrogen Leak in MSIV The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings, for the failure of the licensee to accomplish activities affecting quality in accordance with written procedures. Specifically, operations personnel failed to follow Procedure EN-OP-104, Operability Determinations, and declared main steam isolation valve 1 operable with a through-wall leak on an ASME Class 3 system, despite procedural guidance to the contrary. The licensee has entered this issue into the corrective action program as CR-WF3-2013-1284, and has implemented an ASME Code leak repair as corrective action to restore the degraded condition and reinforced expectations of procedural use and adherence with operations personnel.
The inspectors concluded that the failure of operations personnel to follow procedures was a performance deficiency.
The inspectors determined that the performance deficiency is more than minor, therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its 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 issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined that all questions in Exhibit 2, A. could be answered no, and as such the issue was of very low safety significance (Green).
The inspectors determined this finding has a cross-cutting aspect in the area of human performance associated with the component of decision making because the licensee failed to make conservative assumptions when assessing the source of the nitrogen leak and failed to validate underlying assumptions on subsequent operability reviews [H.1(b)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Perform Testing to Demonstrate Local Manual Operation Action on Safety-Related Air-Operated Valves The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee failed to identify and perform testing on safety-related components to demonstrate that they would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform testing on several safety-related air-operated valves to demonstrate local manual operation in the event their safety-related nitrogen accumulators were exhausted. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service in accordance with requirements Page 5 of 10
 
4Q/2013 Inspection Findings - Waterford 3 contained in the updated final safety analysis report (UFSAR) and design calculations. The licensee entered this issue into their corrective action program as CR-WF3-2012-6703. The immediate corrective actions taken to restore compliance included developing a test for the local manual operation for some valves and the installation of a backup air supply to recharge the accumulators for other valves.
The failure to identify and perform testing to demonstrate that safety-related air-operated valves would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of a system function of safety related equipment. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-8/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included losses of offsite power, which result in an early loss of the instrument air compressors. The fact the accumulators would allow continued air operated valve operation for ten or more hours helped to reduce the risk. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary [P.1(c)].
Inspection Report# : 2013002 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Page 6 of 10
 
4Q/2013 Inspection Findings - Waterford 3 Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering backlogs low enough to support safety [H.2(a)].
Inspection Report# : 2012008 (pdf)
Barrier Integrity Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis Report The inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report, Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193.
The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the Updated Final Safety Analysis Report in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This non-cited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Action 4 of TS 3.3.1 during shutdown with the protective system trip breakers in the open position for Modes 4 and 5 The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that involved reactivity changes use to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not account for temperature increases with a positive MTC within the shutdown margin calculation. This had the potential to affect the available shutdown margin. The licensee entered this Page 7 of 10
 
4Q/2013 Inspection Findings - Waterford 3 issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3-2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service.
The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions
[H.1.a].
Inspection Report# : 2013003 (pdf)
Emergency Preparedness Significance:      Dec 06, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Critique Weaknesses During an Evaluated Exercise The inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to identify deficiencies resulting from the licensees 2013 biennial evaluated exercise. Specifically, the licensee did not identify as part of the critique process two examples of failure to provide a range of protective actions for emergency workers. First, actions were not taken to minimize radiological dose for one in-plant repair team; second, the licensee did not perform habitability evaluations to determine the suitability for continued use of emergency response facilities during the simulated radiological emergency.
The failure to identify weaknesses occurring in an exercise is a performance deficiency. The performance deficiency is more than minor because it is associated with the ERO performance attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continuing these behaviors could result in unnecessary radiological dose to emergency workers and the public in an actual event. Using NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green).
The finding had a cross-cutting aspect in the correction action program component of the problem identification and Page 8 of 10
 
4Q/2013 Inspection Findings - Waterford 3 resolution cross-cutting area because the licensee failed to thoroughly evaluate two issues during the exercise critique process.
Inspection Report# : 2013005 (pdf)
Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
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4Q/2013 Inspection Findings - Waterford 3 Last modified : April 03, 2014 Page 10 of 10
 
1Q/2014 Inspection Findings - Waterford 3 Waterford 3 1Q/2014 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system following steam generator replacement The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide adequate post modification testing instructions for activities affecting quality to the circumstances that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system.
The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring on the feedwater piping system. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety [P.2.b].
Inspection Report# : 2013003 (pdf)
Mitigating Systems Significance:        Feb 20, 2014 Identified By: NRC Item Type: NCV NonCited Violation Page 1 of 11
 
1Q/2014 Inspection Findings - Waterford 3 Failure to Establish Adequate Design Control Measures for the Selection and Review for the Suitability of Application of Molded Case Circuit Breakers (Section 4OA2.2)
A self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, occurred because the licensee did not establish design control measures for the selection and review for the suitability of application of a molded case circuit breaker that was essential to the safety-related function of a shutdown cooling heat exchanger fan cooler. Specifically, the licensee did not select and review for the suitability of the correct safety-related circuit breaker for the application to provide circuit fault protection to the train B shutdown cooling heat exchanger air handling unit fan motor. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-02316 and CR-WF3-2013-04644. The immediate corrective action taken to restore compliance included the replacement of the breaker with a breaker more suitable for the application to protect the air handling unit fan motor. The planned corrective actions included an extent of condition review for other installed breakers and the revision of work order instructions to eliminate the practice of substituting and using the factory acceptance testing for pre-installation and post-maintenance tests, respectively.
The inspectors concluded that the failure to establish design control measures for the selection and review for suitability of application for the correct safety-related circuit breaker was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect breaker affected the availability, reliability, and capability of the shutdown cooling heat exchanger fan coolers to remove heat from the shutdown cooling heat exchanger areas following a design basis accident. The inspectors performed the initial significance determination. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The total exposure period was 23 days. The allowed outage time was 7 days. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was 5E-13/year (Green). The dominant core damage sequences included loss of offsite power events, failure of both trains of containment spray, and the failure of a pressurizer safety relief valve to remain closed. The equipment that helped mitigate the risk included the emergency diesel generators and the essential feedwater systems.
The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect of avoiding complacency in the human performance area because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk on relying on 21 year old vendor information and installing a breaker without pre-installation and adequate post-maintenance testing.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 08, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Replace an Essential Chiller Oil Pump prior to the End of Duty Life.
A Green self-revealing, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish preventative maintenance schedule to inspect or replace an item that have a specific lifetime. Specifically, the licensee did not establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the pump motor exceeding its duty life. As a result, the pump associated with essential chiller B failed in-service. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00095. The immediate corrective action taken to restore compliance was to issue an action request to establish the periodic replacement of the essential chiller pumps prior to the end of their vendor Page 2 of 11
 
1Q/2014 Inspection Findings - Waterford 3 recommended service life.
The failure to establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the end of the vendor provided duty life was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a preventative maintenance schedule to inspect or replace the oil pumps associated with the essential chillers prior to the duty life resulted in the failure of a pump while in service and the unavailability of essential chiller B. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Procedures for Using the Alternate Emergency Fuel Oil Storage Tank Fill Line.
An NRC-identified Green, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish written procedures for filling emergency power sources. Specifically, the licensee did not establish procedures to fill the fuel oil storage tanks for the emergency diesel generators using the safety related, seismic category I alternate emergency fill line. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00636. The immediate corrective action taken to restore compliance was to initiate actions for developing procedures for filling the emergency diesel generator fuel oil storage tanks using the alternate emergency fill line.
The failure to develop procedures for filling emergency power sources was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the failure to establish procedures for the filling of the emergency diesel fuel oil storage tanks using the Seismic Category I alternate emergency fill connection reduced the licensees capability and reliability to for filling the fuel oil storage tanks following an extreme weather event. The inspectors inspector performed the initial significance determination and used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 2, Mitigating Systems Screening Questions, to evaluate this issue.
The finding required a detailed risk evaluation because the performance deficiency could have resulted in a loss of safety function (onsite AC power) because the system may not have remained operable for its 30 day design basis accident mission time. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation for this issue. The analyst determined that the finding was of very low safety significance (Green) because the diesel generators would have remained functional for their 24-hour probabilistic risk assessment mission time. This shorter mission time is used for detailed risk evaluations because, after 24 hours, the NRC assumes that the licensee has substantially more resources available to help mitigate the accident. The dominant core damage sequences included longer term loss of offsite power events and the common cause failure of the diesel generators because of potential problems refilling the diesel fuel oil storage tanks. The relatively long period prior to ultimate diesel generator failure helped to minimize the risk. The finding was not a significant contributor to the large early release frequency. The inspectors concluded that the finding reflected current licensee performance and involved an avoiding complacency cross-cutting aspect of the human performance area in that the licensee did not recognize and plan for the possibility Page 3 of 11
 
1Q/2014 Inspection Findings - Waterford 3 of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 03, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Evaluation for Transient Combustibles.
An NRC-identified Green, non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9 and the Fire Protection Program occurred because the licensee failed to follow procedures.
Specifically, the licensee did not perform a transient combustible evaluation as required by EN-DC-161, Control of Combustibles, to evaluate the impact of capturing and storing up to two gallons of leaking fuel oil in the train B emergency diesel generator room. As a result, the licensee was not performing required hourly fire watches. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-6020 and CR-WF3-2013-06123. The immediate corrective action taken to restore compliance was to perform a transient combustible evaluation implement hourly fire watches.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform a transient combustible evaluation when a flammable liquid above one pint in an approved container was present in the B emergency diesel generator room prevented the licensee from implementing required compensatory measures in response to the presence of transient combustibles. In addition, similar to NRC Inspection Manual Chapter 0612, Appendix E, Section 4, Example k of a more than minor violation, the failure of the leak collection device resulting in fuel oil around emergency diesel generator B represented a credible fire scenario involving transient combustibles that could affect equipment important to safety. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding under Fire Prevention and Administrative Controls and qualitatively screened it as very low safety significance (Green) because the impact of the fire finding was limited to no more than one train of equipment important to safety. The inspectors concluded that the finding reflected current licensee performance and involved a conservative bias cross-cutting aspect in the human performance area in that the licensee did not use decision making practices that emphasized prudent choices over those that are simply allowable.
Inspection Report# : 2014002 (pdf)
Significance:        Dec 20, 2013 Identified By: NRC Item Type: AV Apparent Violation Failure to establish an adequate test program to demonstrate that the train B EDG exhaust fan would perform satisfactorily in service Inspection Report# : 2013008 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Fire Protection Program Procedure Requirements When Securing from a Fire Watch Page 4 of 11
 
1Q/2014 Inspection Findings - Waterford 3 The inspectors identified a non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9, because the licensee did not implement fire protection procedure FP-001-014, Duties of a Fire Watch. Specifically, the licensees fire watch personnel did not implement Section 6.5 of FP-001-014 to remove firefighting equipment from work areas when securing from a fire watch. As a result, multiple undercharged fire extinguishers were left in a fire area. The inspectors determined that this would affect safety-related equipment because it would delay the response to fires in the fire areas. The licensee entered this condition into their corrective action program as CR-WF3-2013-03398 and CR WF3-2013-03523 for resolution. The immediate corrective actions taken to restore compliance included the removal of all undercharged fire extinguishers from deactivated posts and returning them to their proper storage location.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to remove undercharged fire extinguishers from work areas that contained safe shutdown equipment could hinder responses to fires in the fire area. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, to determine the significance of the finding. The inspectors determined that the finding had a low degradation rating because it reflected a fire protection program element whose performance and reliability would be minimally impacted.
Specifically, in all cases identified, there were permanent fully charged portable fire extinguishers of the proper type nearby. Therefore, the finding was of very low safety significance (Green). The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to accomplish activities affecting quality on a degraded safety-related solenoid valve in accordance with procedure requirements The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not accomplish activities affecting quality on a degraded safety-related train B component cooling water (CCW) bypass valve (CC-134B) in accordance with maintenance procedure EN-MA-101, Fundamentals of Maintenance. Specifically, the licensee did not control and perform testing on a leaking solenoid valve related to the operation of a safety-related bypass valve (CC-134B) after maintenance personnel removed the degraded equipment from service as required by Section 5.10 of EN-MA-101. As a result, the licensee could not characterize and determine the cause of the leakage for the safety-related valve. The inspectors determined that this would challenge the safety function of the valve to provide CCW to the ultimate heat sink following a tornado event.
The licensee entered this condition into their corrective action program as CR-WF3-2012-05991, CR-WF3-2012-06288, and CR WF3-2013-04047. The immediate corrective actions taken to restore compliance included the installation of a new valve and debriefing personnel about controlling equipment removed from service when combining preventative and corrective maintenance tasks in one work order.
The failure to control failed equipment removed from the plant to determine the cause in accordance with maintenance procedure requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating Page 5 of 11
 
1Q/2014 Inspection Findings - Waterford 3 events to prevent undesirable consequences. Specifically, the degraded condition challenged the safety function of the valve (CC-134B) to limit the loss of CCW through damaged portions of the dry cooling tower fans following a tornado-generated missile strike. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it was potentially risk significant for an external event (tornado). Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The senior reactor analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 3E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of offsite power, failure of the dry cooling tower pressure boundary, failure to isolate the damaged dry cooling tower, and failure to recover instrument air. The redundant train A component cooling water system combined with the tornado frequency helped to reduce the risk exposure. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work control component of the human performance area in that the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope or activity on plant and human performance [H.3(b)].
Inspection Report# : 2013004 (pdf)
Significance:        Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide adequate design control measures for verifying or checking the adequacy of the effects of a PMP flooding event on the reactor auxiliary building roof areas that contained safety-rea The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on the reactor auxiliary building (RAB) roof areas.
Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012-7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas.
The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1.a].
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1Q/2014 Inspection Findings - Waterford 3 Inspection Report# : 2013003 (pdf)
Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to submit a licensee event report within 60 days of discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corr The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER.
The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency.
However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC's ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:        Sep 24, 2012 Identified By: NRC Item Type: VIO Violation Failure to Take Timely Corrective Action to Establish a Basis for Flood Control Measures The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to establish measures to assure that applicable regulatory requirements and the design basis as defined in 10 CFR 50.2 are correctly translated into procedures. Specifically, the licensee has not determined a basis for the level at which flood control measures are initiated, two years after receiving a non-cited violation for the same deficiency.
As an interim compensatory measure for a previous violation of inadequate technical specifications, the licensee modified their flooding procedure to include an action to start shutting flood control doors at a river level of 24 feet instead of 27 feet. The licensee recognized the need to establish a basis for initiating these actions at 24 feet, and issued a corrective action to track completion. The licensee extended the due date several times and had not completed it by the arrival of the inspection team. The inspection team questioned why the licensee had not completed the calculation to justify their basis for their compensatory measures, noting that it had been over two years since the original violation was identified. The inspectors verified through walk-downs, procedure reviews, and historical data that the licensees use of 24 feet did not represent an immediate operability concern, and that the current river level was sufficiently low to allow time for the licensee to correct the deficiency. This finding was entered into the licensees corrective action program as condition report CR-WF3-2012-03752.
The failure to complete the corrective action to establish a basis for flood control measures in a timely manner was a performance deficiency. The performance deficiency was more than minor because it was associated with the Page 7 of 11
 
1Q/2014 Inspection Findings - Waterford 3 protection from external events 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. Specifically, the licensee failed to verify through calculations or analysis that the actions taken to secure flood doors could be completed in time to protect safety-related equipment from flooding due to a levee failure.
In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the issue was determined to have very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event.
Specifically, the inspectors confirmed that the licensee could reasonably ensure the flood control doors could perform their safety function. This finding had a cross-cutting aspect in the human performance area, resources component in that the licensee failed to maintain long term plant safety by maintenance of design margins and ensuring engineering backlogs low enough to support safety [H.2(a)].
Inspection Report# : 2012008 (pdf)
Barrier Integrity Significance: N/A Jun 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis Report The inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report, Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193.
The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the Updated Final Safety Analysis Report in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This non-cited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
Inspection Report# : 2013003 (pdf)
Significance:      Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to comply with Action 4 of TS 3.3.1 during shutdown with the protective system trip breakers in the open position for Modes 4 and 5 The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that Page 8 of 11
 
1Q/2014 Inspection Findings - Waterford 3 involved reactivity changes use to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not account for temperature increases with a positive MTC within the shutdown margin calculation. This had the potential to affect the available shutdown margin. The licensee entered this issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3-2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service.
The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions
[H.1.a].
Inspection Report# : 2013003 (pdf)
Emergency Preparedness Significance:      Dec 06, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Critique Weaknesses During an Evaluated Exercise The inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to identify deficiencies resulting from the licensees 2013 biennial evaluated exercise. Specifically, the licensee did not identify as part of the critique process two examples of failure to provide a range of protective actions for emergency workers. First, actions were not taken to minimize radiological dose for one in-plant repair team; second, the licensee did not perform habitability evaluations to determine the suitability for continued use of emergency response facilities during the simulated radiological emergency.
The failure to identify weaknesses occurring in an exercise is a performance deficiency. The performance deficiency is more than minor because it is associated with the ERO performance attribute of the emergency preparedness Page 9 of 11
 
1Q/2014 Inspection Findings - Waterford 3 cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continuing these behaviors could result in unnecessary radiological dose to emergency workers and the public in an actual event. Using NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green).
The finding had a cross-cutting aspect in the correction action program component of the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate two issues during the exercise critique process.
Inspection Report# : 2013005 (pdf)
Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been Page 10 of 11
 
1Q/2014 Inspection Findings - Waterford 3 made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
Last modified : May 30, 2014 Page 11 of 11
 
2Q/2014 Inspection Findings - Waterford 3 Waterford 3 2Q/2014 Plant Inspection Findings Initiating Events Significance:      Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Cause and Implement Corrective Actions to Prevent Recurrence for a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for a failure to identify a cause and take corrective actions to prevent recurrence. Specifically, the licensee did not identify a cause or corrective actions to prevent recurrence for a plant trip and equipment failures caused by elevated main feed system vibrations.
The licensee replaced the steam generators at Waterford 3 during refueling outage 18 in late 2012. Upon returning to power operations the licensee experienced elevated vibration levels and related equipment failures on the main feedwater system and emergency feedwater system. The most significant of these failures included a plant trip after a loss of instrument air to the feedwater regulating valve actuator. The licensee determined that the plant trip was a significant event, and initiated a root cause evaluation through its corrective action process. This root cause determination identified a possible cause, which by the licensees program required additional information to confirm or refute. The licensee initiated a proposal to perform modeling of the steam generator flows to provide this information, but later canceled the action. No corrective actions to prevent recurrence were implemented by the licensee. Actions taken to date by the licensee appear to have been effective in mitigating known effects of the vibrations. The licensee documented its failure to determine and document the cause of these vibrations in Condition Report CR-WF3-2014-03238.
The failure to identify the cause of the feedwater vibration-induced problems and to take corrective actions to prevent recurrence as required by 10 CFR Part 50, Appendix B, Criterion XVI is a performance deficiency. The performance deficiency is more than minor because if left uncorrected, it could lead to a more significant safety concern.
Specifically, though individual actions were taken to address failures caused by vibrations, no actions were taken to reduce or eliminate the vibrations themselves. Actions that were taken were not treated as corrective actions to prevent recurrence. A lack of corrective actions to prevent recurrence could leave main feedwater components and other components physically connected to the system such as emergency feedwater susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the team determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the initiating events cornerstone, did not result in a reactor trip and the loss of mitigating equipment needed to transition the plant from the onset of the trip to a stable shutdown condition.
This finding has a resources cross-cutting aspect in the human performance area because leaders did not ensure that procedures used at the time the root cause assessment was performed were adequate to support nuclear safety (H.1).
The procedure used by the licensee allowed a root cause assessment to have an indeterminate root cause and thus no corrective actions to prevent recurrence.
Inspection Report# : 2014008 (pdf)
Significance:      Jun 06, 2014 Identified By: NRC Page 1 of 12
 
2Q/2014 Inspection Findings - Waterford 3 Item Type: FIN Finding Failure to Evaluate Operating Experience as Directed in Station Procedure The team identified a finding for the licensees failure to evaluate industry operating experience as directed in the station operating experience program procedure. Specifically, a vendor supplied Technical Bulletin TB-13-1 Steam Generator and Pressurizer Closure Gasket Replacement Frequency, which recommended that all Westinghouse-designed steam generator and pressurizer closure gaskets be replaced at a prescribed frequency, was not evaluated in accordance with station procedures. This resulted in the licensee failing to take action to periodically replace affected gaskets to preclude degradation of the pressure boundary. The licensee documented this performance deficiency in Condition Report CR-WF3-2014-03229 to determine what further actions were needed.
The failure to evaluate operating experience information as required by licensee procedure EN-OP-100, Operating Experience Program, Revision 20, was a performance deficiency. The performance deficiency is more than minor because if left uncorrected it would have the potential to lead to a more safety-significant concern. Specifically, the failure of the licensee to take any action with regard to the technical bulletin recommendation to replace the steam generator gaskets would allow the gaskets to be installed longer than their useful life. The deterioration of gasket material could result in unplanned transients or shutdowns. The finding is therefore associated with the initiating events cornerstone. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (Green) because it was not an actual degradation that could have resulted in exceeding a reactor system leak rate for a small LOCA; could not have affected other systems used to mitigate a LOCA; did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition; and did not involve a complete or partial loss of a support system that contributes to the likelihood of, or causes, an initiating event and affected mitigation equipment.
This finding has a conservative bias cross-cutting aspect in the human performance area (H.14). Specifically, the licensee assumed that the technical bulletin was not based on actual failures and because steam generators had just been replaced, opted not to take further actions to evaluate or initiate any preventative maintenance to replace gaskets.
Inspection Report# : 2014008 (pdf)
Mitigating Systems Significance:        Jun 06, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedures for Securing Dry Cooling Tower Fans A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III occurred when the licensee did not assure that design basis information was translated into specifications, drawings, procedures, and instructions.
Specifically, after a failure revealed new design basis information regarding the need to place a train of dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet, the licensee failed to incorporate this information into procedures. As a result, the failure recurred. The licensee entered this condition into its corrective action program as Condition Reports CR-WF3-2012-05680 and -06908 and updated procedure OP-006-005, Inverters and Distribution, to incorporate the new design basis information into procedures. The licensee documented its failure to timely update design basis information in Condition Report CR-WF3-2014-02981.
The failure to assure that design basis information was translated into specifications, drawings, procedures, and instructions as required by 10 CFR Part 50, Appendix B, Criterion III was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and Page 2 of 12
 
2Q/2014 Inspection Findings - Waterford 3 capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate design basis information regarding the need to place the dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet into specifications, drawings, procedures, and instructions impacted the capability, availability, and reliability of both trains of dry cooling towers. Using NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because the required number of dry cooling towers in the protected train maintained their operability.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee had not taken effective corrective actions to address an issue in a timely manner commensurate with its safety significance (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Condition Adversely Affecting Flooding Mitigation Design The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. On May 19, 2014, the team identified a significant amount of debris on the floor of one of the main steam isolation valve areas. In a probable maximum precipitation event, this debris could have prevented sufficient water removal by the floor drains to meet design basis assumptions. Following identification, the licensee entered this condition into its corrective action program as Condition Report CR-WF3-2014-03037 and removed the debris from the area.
Excessive debris in the main steam isolation valve A area that could challenge the waterremoval capability of safety-related drain systems was a condition adverse to quality. The licensees failure to promptly identify and correct this condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. The lead inspector performed the initial significance determination for performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because it involved the degradation of equipment specifically designed to mitigate a flooding event. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The bounding change to the core damage frequency was 4.7x10-8 per year (Green). The dominant core damage sequences included extremely heavy rainfall, a loss of offsite power initiating event, failure of the train B 4.16kV bus, and failure of the pressurizer safety relief valves to close. The low initiating event frequency reduced the risk significance.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensees corrective actions from the previous non-cited violation did not fully address the issue (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Multiple Degraded or Nonconforming Conditions The inspectors identified multiple instances of the licensees failure to promptly correct degraded or nonconforming conditions as required by 10 CFR Part 50, Appendix B, Criterion XVI. At the conclusion of the inspection, the Page 3 of 12
 
2Q/2014 Inspection Findings - Waterford 3 licensee had one structure, system or component that had been degraded since November 2008, requiring compensatory measures to provide reasonable assurance of operability; the licensee had another degraded condition that had existed since April 2011 with no compensatory measures in place. Following the teams identification of this issue, the licensee documented this issue in Condition Report CR-WF3-2014-03250 to evaluate the timeliness of its corrective actions.
The failure to promptly correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency is more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, the team determined this finding to be of very low safety significance (Green) because it did not represent the actual loss of function of a safety-related system or train.
This finding has an evaluation cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate the issues to ensure that the resolutions addressed causes and extents of condition commensurate with the issues safety significance (P.2).
Inspection Report# : 2014008 (pdf)
Significance:        Feb 20, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Adequate Design Control Measures for the Selection and Review for the Suitability of Application of Molded Case Circuit Breakers A self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, occurred because the licensee did not establish design control measures for the selection and review for the suitability of application of a molded case circuit breaker that was essential to the safety-related function of a shutdown cooling heat exchanger fan cooler. Specifically, the licensee did not select and review for the suitability of the correct safety-related circuit breaker for the application to provide circuit fault protection to the train B shutdown cooling heat exchanger air handling unit fan motor. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-02316 and CR-WF3-2013-04644. The immediate corrective action taken to restore compliance included the replacement of the breaker with a breaker more suitable for the application to protect the air handling unit fan motor. The planned corrective actions included an extent of condition review for other installed breakers and the revision of work order instructions to eliminate the practice of substituting and using the factory acceptance testing for pre-installation and post-maintenance tests, respectively.
The inspectors concluded that the failure to establish design control measures for the selection and review for suitability of application for the correct safety-related circuit breaker was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect breaker affected the availability, reliability, and capability of the shutdown cooling heat exchanger fan coolers to remove heat from the shutdown cooling heat exchanger areas following a design basis accident. The inspectors performed the initial significance determination. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The total exposure period was 23 Page 4 of 12
 
2Q/2014 Inspection Findings - Waterford 3 days. The allowed outage time was 7 days. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was 5E-13/year (Green). The dominant core damage sequences included loss of offsite power events, failure of both trains of containment spray, and the failure of a pressurizer safety relief valve to remain closed. The equipment that helped mitigate the risk included the emergency diesel generators and the essential feedwater systems.
The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect of avoiding complacency in the human performance area because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk on relying on 21 year old vendor information and installing a breaker without pre-installation and adequate post-maintenance testing.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 08, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Replace an Essential Chiller Oil Pump prior to the End of Duty Life.
A Green self-revealing, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish preventative maintenance schedule to inspect or replace an item that have a specific lifetime. Specifically, the licensee did not establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the pump motor exceeding its duty life. As a result, the pump associated with essential chiller B failed in-service. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00095. The immediate corrective action taken to restore compliance was to issue an action request to establish the periodic replacement of the essential chiller pumps prior to the end of their vendor recommended service life.
The failure to establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the end of the vendor provided duty life was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a preventative maintenance schedule to inspect or replace the oil pumps associated with the essential chillers prior to the duty life resulted in the failure of a pump while in service and the unavailability of essential chiller B. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Procedures for Using the Alternate Emergency Fuel Oil Storage Tank Fill Line.
An NRC-identified Green, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish written procedures for filling emergency power sources. Specifically, the licensee did not establish procedures to fill the fuel oil storage tanks for the emergency diesel generators using the safety related, seismic category I alternate emergency fill line. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00636. The immediate corrective action taken to restore compliance was to initiate actions for developing procedures for filling the emergency diesel generator fuel oil storage tanks using the alternate Page 5 of 12
 
2Q/2014 Inspection Findings - Waterford 3 emergency fill line.
The failure to develop procedures for filling emergency power sources was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the failure to establish procedures for the filling of the emergency diesel fuel oil storage tanks using the Seismic Category I alternate emergency fill connection reduced the licensees capability and reliability to for filling the fuel oil storage tanks following an extreme weather event. The inspectors inspector performed the initial significance determination and used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 2, Mitigating Systems Screening Questions, to evaluate this issue.
The finding required a detailed risk evaluation because the performance deficiency could have resulted in a loss of safety function (onsite AC power) because the system may not have remained operable for its 30 day design basis accident mission time. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation for this issue. The analyst determined that the finding was of very low safety significance (Green) because the diesel generators would have remained functional for their 24-hour probabilistic risk assessment mission time. This shorter mission time is used for detailed risk evaluations because, after 24 hours, the NRC assumes that the licensee has substantially more resources available to help mitigate the accident. The dominant core damage sequences included longer term loss of offsite power events and the common cause failure of the diesel generators because of potential problems refilling the diesel fuel oil storage tanks. The relatively long period prior to ultimate diesel generator failure helped to minimize the risk. The finding was not a significant contributor to the large early release frequency. The inspectors concluded that the finding reflected current licensee performance and involved an avoiding complacency cross-cutting aspect of the human performance area in that the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 03, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Evaluation for Transient Combustibles.
An NRC-identified Green, non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9 and the Fire Protection Program occurred because the licensee failed to follow procedures.
Specifically, the licensee did not perform a transient combustible evaluation as required by EN-DC-161, Control of Combustibles, to evaluate the impact of capturing and storing up to two gallons of leaking fuel oil in the train B emergency diesel generator room. As a result, the licensee was not performing required hourly fire watches. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-6020 and CR-WF3-2013-06123. The immediate corrective action taken to restore compliance was to perform a transient combustible evaluation implement hourly fire watches.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform a transient combustible evaluation when a flammable liquid above one pint in an approved container was present in the B emergency diesel generator room prevented the licensee from implementing required compensatory measures in response to the presence of transient combustibles. In addition, similar to NRC Inspection Manual Chapter 0612, Appendix E, Section 4, Example k of a more than minor violation, the failure of the leak collection device resulting in fuel oil around emergency diesel generator B represented a credible fire scenario involving transient combustibles that could affect equipment important to safety. The inspectors Page 6 of 12
 
2Q/2014 Inspection Findings - Waterford 3 used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding under Fire Prevention and Administrative Controls and qualitatively screened it as very low safety significance (Green) because the impact of the fire finding was limited to no more than one train of equipment important to safety. The inspectors concluded that the finding reflected current licensee performance and involved a conservative bias cross-cutting aspect in the human performance area in that the licensee did not use decision making practices that emphasized prudent choices over those that are simply allowable.
Inspection Report# : 2014002 (pdf)
Significance:        Dec 20, 2013 Identified By: NRC Item Type: VIO Violation Failure to establish an adequate test program to demonstrate that the train B EDG exhaust fan would perform satisfactorily in service A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criteria XI, Test Control, occurred because the licensee failed to establish an adequate test program to demonstrate that a safety related component associated with the Train B Emergency Diesel Generator would perform satisfactorily in service. Specifically, the licensee failed to identify and perform adequate testing on the Train B EDG exhaust fan to demonstrate that the exhaust fan would perform satisfactorily in service, which incorporated the requirements and acceptance limits contained in applicable design documents such as the Final Safety Analysis Report, as updated. As a result, the licensee failed to ensure that for all operational tests that the safety related exhaust fan would perform satisfactorily such that it would provide sufficient flow and remove heat during accident conditions. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-02530. The immediate corrective actions taken to restore compliance included the replacement of the B EDG exhaust fan assembly. The planned corrective actions include the review of the EDG ventilation system monitoring plan.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the licensee failed to perform testing to ensure that the B EDG ventilation exhaust fan would fulfill its safety function to remove heat from the EDG room when the diesel operates during accident conditions. The senior resident inspector performed the initial significance determination for the diesel generator room ventilation fan failure.
The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. The emergency diesel generator needed the ventilation exhaust fan to remain Operable. The unit was not recoverable. The total exposure period was 25 days. The allowed outage time was 72 hours. The analyst determined the best estimated change to the core damage frequency was 4.4E-6/year (White). The risk significance was low to moderate (White). The dominant core damage sequences included loss of offsite power events, leading to station blackout, coincident with the failure of the turbine driven auxiliary feedwater pump. Equipment that helped mitigated the risk included recovery of an emergency diesel generator or manually starting a temporary emergency diesel generator set. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the resource component of the human performance area in that the licensee did not have complete, accurate and up-to-date operational surveillance procedure tests [H.2.c].
Inspection Report# : 2013008 (pdf)
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2Q/2014 Inspection Findings - Waterford 3 Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Fire Protection Program Procedure Requirements When Securing from a Fire Watch The inspectors identified a non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9, because the licensee did not implement fire protection procedure FP-001-014, Duties of a Fire Watch. Specifically, the licensees fire watch personnel did not implement Section 6.5 of FP-001-014 to remove firefighting equipment from work areas when securing from a fire watch. As a result, multiple undercharged fire extinguishers were left in a fire area. The inspectors determined that this would affect safety-related equipment because it would delay the response to fires in the fire areas. The licensee entered this condition into their corrective action program as CR-WF3-2013-03398 and CR WF3-2013-03523 for resolution. The immediate corrective actions taken to restore compliance included the removal of all undercharged fire extinguishers from deactivated posts and returning them to their proper storage location.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to remove undercharged fire extinguishers from work areas that contained safe shutdown equipment could hinder responses to fires in the fire area. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, to determine the significance of the finding. The inspectors determined that the finding had a low degradation rating because it reflected a fire protection program element whose performance and reliability would be minimally impacted.
Specifically, in all cases identified, there were permanent fully charged portable fire extinguishers of the proper type nearby. Therefore, the finding was of very low safety significance (Green). The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2013004 (pdf)
Significance:        Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to accomplish activities affecting quality on a degraded safety-related solenoid valve in accordance with procedure requirements The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not accomplish activities affecting quality on a degraded safety-related train B component cooling water (CCW) bypass valve (CC-134B) in accordance with maintenance procedure EN-MA-101, Fundamentals of Maintenance. Specifically, the licensee did not control and perform testing on a leaking solenoid valve related to the operation of a safety-related bypass valve (CC-134B) after maintenance personnel removed the degraded equipment from service as required by Section 5.10 of EN-MA-101. As a result, the licensee could not characterize and determine the cause of the leakage for the safety-related valve. The inspectors determined that this would challenge the safety function of the valve to provide CCW to the ultimate heat sink following a tornado event.
The licensee entered this condition into their corrective action program as CR-WF3-2012-05991, CR-WF3-2012-06288, and CR WF3-2013-04047. The immediate corrective actions taken to restore compliance included the installation of a new valve and debriefing personnel about controlling equipment removed from service when combining preventative and corrective maintenance tasks in one work order.
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2Q/2014 Inspection Findings - Waterford 3 The failure to control failed equipment removed from the plant to determine the cause in accordance with maintenance procedure requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition challenged the safety function of the valve (CC-134B) to limit the loss of CCW through damaged portions of the dry cooling tower fans following a tornado-generated missile strike. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it was potentially risk significant for an external event (tornado). Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The senior reactor analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 3E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of offsite power, failure of the dry cooling tower pressure boundary, failure to isolate the damaged dry cooling tower, and failure to recover instrument air. The redundant train A component cooling water system combined with the tornado frequency helped to reduce the risk exposure. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work control component of the human performance area in that the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope or activity on plant and human performance [H.3(b)].
Inspection Report# : 2013004 (pdf)
Barrier Integrity Emergency Preparedness Significance:        Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions.
The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860.
The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness Page 9 of 12
 
2Q/2014 Inspection Findings - Waterford 3 cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes.
Inspection Report# : 2014003 (pdf)
Significance:        Dec 06, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Critique Weaknesses During an Evaluated Exercise The inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to identify deficiencies resulting from the licensees 2013 biennial evaluated exercise. Specifically, the licensee did not identify as part of the critique process two examples of failure to provide a range of protective actions for emergency workers. First, actions were not taken to minimize radiological dose for one in-plant repair team; second, the licensee did not perform habitability evaluations to determine the suitability for continued use of emergency response facilities during the simulated radiological emergency.
The failure to identify weaknesses occurring in an exercise is a performance deficiency. The performance deficiency is more than minor because it is associated with the ERO performance attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continuing these behaviors could result in unnecessary radiological dose to emergency workers and the public in an actual event. Using NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green).
The finding had a cross-cutting aspect in the correction action program component of the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate two issues during the exercise critique process.
Inspection Report# : 2013005 (pdf)
Occupational Radiation Safety Significance:        Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Entry into a High Radiation Area The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker Page 10 of 12
 
2Q/2014 Inspection Findings - Waterford 3 entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638.
The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014003 (pdf)
Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety Page 11 of 12
 
2Q/2014 Inspection Findings - Waterford 3 significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
Inspection Report# : 2014008 (pdf)
Last modified : August 29, 2014 Page 12 of 12
 
3Q/2014 Inspection Findings - Waterford 3 Waterford 3 3Q/2014 Plant Inspection Findings Initiating Events Significance:      Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Cause and Implement Corrective Actions to Prevent Recurrence for a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for a failure to identify a cause and take corrective actions to prevent recurrence. Specifically, the licensee did not identify a cause or corrective actions to prevent recurrence for a plant trip and equipment failures caused by elevated main feed system vibrations.
The licensee replaced the steam generators at Waterford 3 during refueling outage 18 in late 2012. Upon returning to power operations the licensee experienced elevated vibration levels and related equipment failures on the main feedwater system and emergency feedwater system. The most significant of these failures included a plant trip after a loss of instrument air to the feedwater regulating valve actuator. The licensee determined that the plant trip was a significant event, and initiated a root cause evaluation through its corrective action process. This root cause determination identified a possible cause, which by the licensees program required additional information to confirm or refute. The licensee initiated a proposal to perform modeling of the steam generator flows to provide this information, but later canceled the action. No corrective actions to prevent recurrence were implemented by the licensee. Actions taken to date by the licensee appear to have been effective in mitigating known effects of the vibrations. The licensee documented its failure to determine and document the cause of these vibrations in Condition Report CR-WF3-2014-03238.
The failure to identify the cause of the feedwater vibration-induced problems and to take corrective actions to prevent recurrence as required by 10 CFR Part 50, Appendix B, Criterion XVI is a performance deficiency. The performance deficiency is more than minor because if left uncorrected, it could lead to a more significant safety concern.
Specifically, though individual actions were taken to address failures caused by vibrations, no actions were taken to reduce or eliminate the vibrations themselves. Actions that were taken were not treated as corrective actions to prevent recurrence. A lack of corrective actions to prevent recurrence could leave main feedwater components and other components physically connected to the system such as emergency feedwater susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the team determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the initiating events cornerstone, did not result in a reactor trip and the loss of mitigating equipment needed to transition the plant from the onset of the trip to a stable shutdown condition.
This finding has a resources cross-cutting aspect in the human performance area because leaders did not ensure that procedures used at the time the root cause assessment was performed were adequate to support nuclear safety (H.1).
The procedure used by the licensee allowed a root cause assessment to have an indeterminate root cause and thus no corrective actions to prevent recurrence.
Inspection Report# : 2014008 (pdf)
Significance:      Jun 06, 2014 Identified By: NRC Page 1 of 10
 
3Q/2014 Inspection Findings - Waterford 3 Item Type: FIN Finding Failure to Evaluate Operating Experience as Directed in Station Procedure The team identified a finding for the licensees failure to evaluate industry operating experience as directed in the station operating experience program procedure. Specifically, a vendor supplied Technical Bulletin TB-13-1 Steam Generator and Pressurizer Closure Gasket Replacement Frequency, which recommended that all Westinghouse-designed steam generator and pressurizer closure gaskets be replaced at a prescribed frequency, was not evaluated in accordance with station procedures. This resulted in the licensee failing to take action to periodically replace affected gaskets to preclude degradation of the pressure boundary. The licensee documented this performance deficiency in Condition Report CR-WF3-2014-03229 to determine what further actions were needed.
The failure to evaluate operating experience information as required by licensee procedure EN-OP-100, Operating Experience Program, Revision 20, was a performance deficiency. The performance deficiency is more than minor because if left uncorrected it would have the potential to lead to a more safety-significant concern. Specifically, the failure of the licensee to take any action with regard to the technical bulletin recommendation to replace the steam generator gaskets would allow the gaskets to be installed longer than their useful life. The deterioration of gasket material could result in unplanned transients or shutdowns. The finding is therefore associated with the initiating events cornerstone. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (Green) because it was not an actual degradation that could have resulted in exceeding a reactor system leak rate for a small LOCA; could not have affected other systems used to mitigate a LOCA; did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition; and did not involve a complete or partial loss of a support system that contributes to the likelihood of, or causes, an initiating event and affected mitigation equipment.
This finding has a conservative bias cross-cutting aspect in the human performance area (H.14). Specifically, the licensee assumed that the technical bulletin was not based on actual failures and because steam generators had just been replaced, opted not to take further actions to evaluate or initiate any preventative maintenance to replace gaskets.
Inspection Report# : 2014008 (pdf)
Mitigating Systems Significance:        Jun 06, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedures for Securing Dry Cooling Tower Fans A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III occurred when the licensee did not assure that design basis information was translated into specifications, drawings, procedures, and instructions.
Specifically, after a failure revealed new design basis information regarding the need to place a train of dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet, the licensee failed to incorporate this information into procedures. As a result, the failure recurred. The licensee entered this condition into its corrective action program as Condition Reports CR-WF3-2012-05680 and -06908 and updated procedure OP-006-005, Inverters and Distribution, to incorporate the new design basis information into procedures. The licensee documented its failure to timely update design basis information in Condition Report CR-WF3-2014-02981.
The failure to assure that design basis information was translated into specifications, drawings, procedures, and instructions as required by 10 CFR Part 50, Appendix B, Criterion III was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and Page 2 of 10
 
3Q/2014 Inspection Findings - Waterford 3 capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate design basis information regarding the need to place the dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet into specifications, drawings, procedures, and instructions impacted the capability, availability, and reliability of both trains of dry cooling towers. Using NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because the required number of dry cooling towers in the protected train maintained their operability.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee had not taken effective corrective actions to address an issue in a timely manner commensurate with its safety significance (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Condition Adversely Affecting Flooding Mitigation Design The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. On May 19, 2014, the team identified a significant amount of debris on the floor of one of the main steam isolation valve areas. In a probable maximum precipitation event, this debris could have prevented sufficient water removal by the floor drains to meet design basis assumptions. Following identification, the licensee entered this condition into its corrective action program as Condition Report CR-WF3-2014-03037 and removed the debris from the area.
Excessive debris in the main steam isolation valve A area that could challenge the waterremoval capability of safety-related drain systems was a condition adverse to quality. The licensees failure to promptly identify and correct this condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. The lead inspector performed the initial significance determination for performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because it involved the degradation of equipment specifically designed to mitigate a flooding event. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The bounding change to the core damage frequency was 4.7x10-8 per year (Green). The dominant core damage sequences included extremely heavy rainfall, a loss of offsite power initiating event, failure of the train B 4.16kV bus, and failure of the pressurizer safety relief valves to close. The low initiating event frequency reduced the risk significance.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensees corrective actions from the previous non-cited violation did not fully address the issue (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Multiple Degraded or Nonconforming Conditions The inspectors identified multiple instances of the licensees failure to promptly correct degraded or nonconforming conditions as required by 10 CFR Part 50, Appendix B, Criterion XVI. At the conclusion of the inspection, the Page 3 of 10
 
3Q/2014 Inspection Findings - Waterford 3 licensee had one structure, system or component that had been degraded since November 2008, requiring compensatory measures to provide reasonable assurance of operability; the licensee had another degraded condition that had existed since April 2011 with no compensatory measures in place. Following the teams identification of this issue, the licensee documented this issue in Condition Report CR-WF3-2014-03250 to evaluate the timeliness of its corrective actions.
The failure to promptly correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency is more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, the team determined this finding to be of very low safety significance (Green) because it did not represent the actual loss of function of a safety-related system or train.
This finding has an evaluation cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate the issues to ensure that the resolutions addressed causes and extents of condition commensurate with the issues safety significance (P.2).
Inspection Report# : 2014008 (pdf)
Significance:        Feb 20, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Adequate Design Control Measures for the Selection and Review for the Suitability of Application of Molded Case Circuit Breakers A self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, occurred because the licensee did not establish design control measures for the selection and review for the suitability of application of a molded case circuit breaker that was essential to the safety-related function of a shutdown cooling heat exchanger fan cooler. Specifically, the licensee did not select and review for the suitability of the correct safety-related circuit breaker for the application to provide circuit fault protection to the train B shutdown cooling heat exchanger air handling unit fan motor. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-02316 and CR-WF3-2013-04644. The immediate corrective action taken to restore compliance included the replacement of the breaker with a breaker more suitable for the application to protect the air handling unit fan motor. The planned corrective actions included an extent of condition review for other installed breakers and the revision of work order instructions to eliminate the practice of substituting and using the factory acceptance testing for pre-installation and post-maintenance tests, respectively.
The inspectors concluded that the failure to establish design control measures for the selection and review for suitability of application for the correct safety-related circuit breaker was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect breaker affected the availability, reliability, and capability of the shutdown cooling heat exchanger fan coolers to remove heat from the shutdown cooling heat exchanger areas following a design basis accident. The inspectors performed the initial significance determination. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The total exposure period was 23 Page 4 of 10
 
3Q/2014 Inspection Findings - Waterford 3 days. The allowed outage time was 7 days. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was 5E-13/year (Green). The dominant core damage sequences included loss of offsite power events, failure of both trains of containment spray, and the failure of a pressurizer safety relief valve to remain closed. The equipment that helped mitigate the risk included the emergency diesel generators and the essential feedwater systems.
The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect of avoiding complacency in the human performance area because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk on relying on 21 year old vendor information and installing a breaker without pre-installation and adequate post-maintenance testing.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 08, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Replace an Essential Chiller Oil Pump prior to the End of Duty Life.
A Green self-revealing, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish preventative maintenance schedule to inspect or replace an item that have a specific lifetime. Specifically, the licensee did not establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the pump motor exceeding its duty life. As a result, the pump associated with essential chiller B failed in-service. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00095. The immediate corrective action taken to restore compliance was to issue an action request to establish the periodic replacement of the essential chiller pumps prior to the end of their vendor recommended service life.
The failure to establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the end of the vendor provided duty life was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a preventative maintenance schedule to inspect or replace the oil pumps associated with the essential chillers prior to the duty life resulted in the failure of a pump while in service and the unavailability of essential chiller B. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Procedures for Using the Alternate Emergency Fuel Oil Storage Tank Fill Line.
An NRC-identified Green, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish written procedures for filling emergency power sources. Specifically, the licensee did not establish procedures to fill the fuel oil storage tanks for the emergency diesel generators using the safety related, seismic category I alternate emergency fill line. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00636. The immediate corrective action taken to restore compliance was to initiate actions for developing procedures for filling the emergency diesel generator fuel oil storage tanks using the alternate Page 5 of 10
 
3Q/2014 Inspection Findings - Waterford 3 emergency fill line.
The failure to develop procedures for filling emergency power sources was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the failure to establish procedures for the filling of the emergency diesel fuel oil storage tanks using the Seismic Category I alternate emergency fill connection reduced the licensees capability and reliability to for filling the fuel oil storage tanks following an extreme weather event. The inspectors inspector performed the initial significance determination and used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 2, Mitigating Systems Screening Questions, to evaluate this issue.
The finding required a detailed risk evaluation because the performance deficiency could have resulted in a loss of safety function (onsite AC power) because the system may not have remained operable for its 30 day design basis accident mission time. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation for this issue. The analyst determined that the finding was of very low safety significance (Green) because the diesel generators would have remained functional for their 24-hour probabilistic risk assessment mission time. This shorter mission time is used for detailed risk evaluations because, after 24 hours, the NRC assumes that the licensee has substantially more resources available to help mitigate the accident. The dominant core damage sequences included longer term loss of offsite power events and the common cause failure of the diesel generators because of potential problems refilling the diesel fuel oil storage tanks. The relatively long period prior to ultimate diesel generator failure helped to minimize the risk. The finding was not a significant contributor to the large early release frequency. The inspectors concluded that the finding reflected current licensee performance and involved an avoiding complacency cross-cutting aspect of the human performance area in that the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 03, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Evaluation for Transient Combustibles.
An NRC-identified Green, non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9 and the Fire Protection Program occurred because the licensee failed to follow procedures.
Specifically, the licensee did not perform a transient combustible evaluation as required by EN-DC-161, Control of Combustibles, to evaluate the impact of capturing and storing up to two gallons of leaking fuel oil in the train B emergency diesel generator room. As a result, the licensee was not performing required hourly fire watches. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-6020 and CR-WF3-2013-06123. The immediate corrective action taken to restore compliance was to perform a transient combustible evaluation implement hourly fire watches.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform a transient combustible evaluation when a flammable liquid above one pint in an approved container was present in the B emergency diesel generator room prevented the licensee from implementing required compensatory measures in response to the presence of transient combustibles. In addition, similar to NRC Inspection Manual Chapter 0612, Appendix E, Section 4, Example k of a more than minor violation, the failure of the leak collection device resulting in fuel oil around emergency diesel generator B represented a credible fire scenario involving transient combustibles that could affect equipment important to safety. The inspectors Page 6 of 10
 
3Q/2014 Inspection Findings - Waterford 3 used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding under Fire Prevention and Administrative Controls and qualitatively screened it as very low safety significance (Green) because the impact of the fire finding was limited to no more than one train of equipment important to safety. The inspectors concluded that the finding reflected current licensee performance and involved a conservative bias cross-cutting aspect in the human performance area in that the licensee did not use decision making practices that emphasized prudent choices over those that are simply allowable.
Inspection Report# : 2014002 (pdf)
Significance:        Dec 20, 2013 Identified By: NRC Item Type: VIO Violation Failure to establish an adequate test program to demonstrate that the train B EDG exhaust fan would perform satisfactorily in service A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criteria XI, Test Control, occurred because the licensee failed to establish an adequate test program to demonstrate that a safety related component associated with the Train B Emergency Diesel Generator would perform satisfactorily in service. Specifically, the licensee failed to identify and perform adequate testing on the Train B EDG exhaust fan to demonstrate that the exhaust fan would perform satisfactorily in service, which incorporated the requirements and acceptance limits contained in applicable design documents such as the Final Safety Analysis Report, as updated. As a result, the licensee failed to ensure that for all operational tests that the safety related exhaust fan would perform satisfactorily such that it would provide sufficient flow and remove heat during accident conditions. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-02530. The immediate corrective actions taken to restore compliance included the replacement of the B EDG exhaust fan assembly. The planned corrective actions include the review of the EDG ventilation system monitoring plan.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the licensee failed to perform testing to ensure that the B EDG ventilation exhaust fan would fulfill its safety function to remove heat from the EDG room when the diesel operates during accident conditions. The senior resident inspector performed the initial significance determination for the diesel generator room ventilation fan failure.
The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. The emergency diesel generator needed the ventilation exhaust fan to remain Operable. The unit was not recoverable. The total exposure period was 25 days. The allowed outage time was 72 hours. The analyst determined the best estimated change to the core damage frequency was 4.4E-6/year (White). The risk significance was low to moderate (White). The dominant core damage sequences included loss of offsite power events, leading to station blackout, coincident with the failure of the turbine driven auxiliary feedwater pump. Equipment that helped mitigated the risk included recovery of an emergency diesel generator or manually starting a temporary emergency diesel generator set. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the resource component of the human performance area in that the licensee did not have complete, accurate and up-to-date operational surveillance procedure tests [H.2.c].
Final significance determination and White NOV issued March 28, 2014. IR 05000382/2014009 (ML14086A768)
Inspection Report# : 2013008 (pdf)
Inspection Report# : 2014009 (pdf)
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3Q/2014 Inspection Findings - Waterford 3 Barrier Integrity Emergency Preparedness Significance:        Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions.
The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860.
The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes.
Inspection Report# : 2014003 (pdf)
Significance:        Dec 06, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Critique Weaknesses During an Evaluated Exercise The inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to identify deficiencies Page 8 of 10
 
3Q/2014 Inspection Findings - Waterford 3 resulting from the licensees 2013 biennial evaluated exercise. Specifically, the licensee did not identify as part of the critique process two examples of failure to provide a range of protective actions for emergency workers. First, actions were not taken to minimize radiological dose for one in-plant repair team; second, the licensee did not perform habitability evaluations to determine the suitability for continued use of emergency response facilities during the simulated radiological emergency.
The failure to identify weaknesses occurring in an exercise is a performance deficiency. The performance deficiency is more than minor because it is associated with the ERO performance attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continuing these behaviors could result in unnecessary radiological dose to emergency workers and the public in an actual event. Using NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green).
The finding had a cross-cutting aspect in the correction action program component of the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate two issues during the exercise critique process.
Inspection Report# : 2013005 (pdf)
Occupational Radiation Safety Significance:      Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Entry into a High Radiation Area The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638.
The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014003 (pdf)
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3Q/2014 Inspection Findings - Waterford 3 Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
Inspection Report# : 2014008 (pdf)
Last modified : November 26, 2014 Page 10 of 10
 
4Q/2014 Inspection Findings - Waterford 3 Waterford 3 4Q/2014 Plant Inspection Findings Initiating Events Significance:      Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Cause and Implement Corrective Actions to Prevent Recurrence for a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for a failure to identify a cause and take corrective actions to prevent recurrence. Specifically, the licensee did not identify a cause or corrective actions to prevent recurrence for a plant trip and equipment failures caused by elevated main feed system vibrations.
The licensee replaced the steam generators at Waterford 3 during refueling outage 18 in late 2012. Upon returning to power operations the licensee experienced elevated vibration levels and related equipment failures on the main feedwater system and emergency feedwater system. The most significant of these failures included a plant trip after a loss of instrument air to the feedwater regulating valve actuator. The licensee determined that the plant trip was a significant event, and initiated a root cause evaluation through its corrective action process. This root cause determination identified a possible cause, which by the licensees program required additional information to confirm or refute. The licensee initiated a proposal to perform modeling of the steam generator flows to provide this information, but later canceled the action. No corrective actions to prevent recurrence were implemented by the licensee. Actions taken to date by the licensee appear to have been effective in mitigating known effects of the vibrations. The licensee documented its failure to determine and document the cause of these vibrations in Condition Report CR-WF3-2014-03238.
The failure to identify the cause of the feedwater vibration-induced problems and to take corrective actions to prevent recurrence as required by 10 CFR Part 50, Appendix B, Criterion XVI is a performance deficiency. The performance deficiency is more than minor because if left uncorrected, it could lead to a more significant safety concern.
Specifically, though individual actions were taken to address failures caused by vibrations, no actions were taken to reduce or eliminate the vibrations themselves. Actions that were taken were not treated as corrective actions to prevent recurrence. A lack of corrective actions to prevent recurrence could leave main feedwater components and other components physically connected to the system such as emergency feedwater susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the team determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the initiating events cornerstone, did not result in a reactor trip and the loss of mitigating equipment needed to transition the plant from the onset of the trip to a stable shutdown condition.
This finding has a resources cross-cutting aspect in the human performance area because leaders did not ensure that procedures used at the time the root cause assessment was performed were adequate to support nuclear safety (H.1).
The procedure used by the licensee allowed a root cause assessment to have an indeterminate root cause and thus no corrective actions to prevent recurrence.
Inspection Report# : 2014008 (pdf)
Significance:      Jun 06, 2014 Identified By: NRC Page 1 of 13
 
4Q/2014 Inspection Findings - Waterford 3 Item Type: FIN Finding Failure to Evaluate Operating Experience as Directed in Station Procedure The team identified a finding for the licensees failure to evaluate industry operating experience as directed in the station operating experience program procedure. Specifically, a vendor supplied Technical Bulletin TB-13-1 Steam Generator and Pressurizer Closure Gasket Replacement Frequency, which recommended that all Westinghouse-designed steam generator and pressurizer closure gaskets be replaced at a prescribed frequency, was not evaluated in accordance with station procedures. This resulted in the licensee failing to take action to periodically replace affected gaskets to preclude degradation of the pressure boundary. The licensee documented this performance deficiency in Condition Report CR-WF3-2014-03229 to determine what further actions were needed.
The failure to evaluate operating experience information as required by licensee procedure EN-OP-100, Operating Experience Program, Revision 20, was a performance deficiency. The performance deficiency is more than minor because if left uncorrected it would have the potential to lead to a more safety-significant concern. Specifically, the failure of the licensee to take any action with regard to the technical bulletin recommendation to replace the steam generator gaskets would allow the gaskets to be installed longer than their useful life. The deterioration of gasket material could result in unplanned transients or shutdowns. The finding is therefore associated with the initiating events cornerstone. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (Green) because it was not an actual degradation that could have resulted in exceeding a reactor system leak rate for a small LOCA; could not have affected other systems used to mitigate a LOCA; did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition; and did not involve a complete or partial loss of a support system that contributes to the likelihood of, or causes, an initiating event and affected mitigation equipment.
This finding has a conservative bias cross-cutting aspect in the human performance area (H.14). Specifically, the licensee assumed that the technical bulletin was not based on actual failures and because steam generators had just been replaced, opted not to take further actions to evaluate or initiate any preventative maintenance to replace gaskets.
Inspection Report# : 2014008 (pdf)
Mitigating Systems Significance:        Dec 31, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Control Potential Tornado-Borne Missile Hazards The inspectors identified a non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 312, on two separate instances. Specifically, on both November 16 and December 23, 2014, the licensee entered the off-normal procedure due to a tornado watch but failed to assess and control potential tornado-borne missile hazards on site as required by the procedure. The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-05912 and CR-WF3-2014-06453. The immediate corrective action taken to restore compliance was to secure the identified hazards.
This finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, in the event of a tornado at the site, these loose items could have become missiles with the potential to impact safety-related site equipment and personnel. The inspectors determined the finding was of very low safety Page 2 of 13
 
4Q/2014 Inspection Findings - Waterford 3 significance (Green) because the it did not involve the loss or degradation of equipment or functions specifically designed to mitigate a seismic, flooding, or severe weather event (e.g. seismic snubbers, flooding barriers, tornado doors). The inspectors concluded that the finding had a cross-cutting aspect in the area of Human Performance, Field Presence, because the licensee did not ensure supervisory and management oversight of work activities.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Operability Determination Process TThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to assess immediate operability of safety-related systems in accordance with site procedures, in three separate instances. Specifically, on two occasions, the licensee did not properly assess operability of safety-related relays in the Engineered Safety Features Actuation Signal system, which in turn brought into question the operability of the emergency diesel generators. A third example involved the licensees failure to properly assess operability of safety-related class 3 piping on the dry cooling towers, which brought into question the operability of the component cooling water system. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-06014. The licensee restored compliance by revising the immediate operability determinations to reflect an adequate reason to justify operability of the systems in questions.
The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to follow the Operability Determination procedure caused the licensee to incorrectly assess the capability of the systems impacted by the relays and dry cooling tower tube leak to perform their safety function and there was a reasonable doubt on the operability of the systems. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process, because individuals did not use a consistent, systematic approach to make a decision and risk insights were not incorporated appropriately.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish an Inspection Schedule of the Dry Cooling Towers The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for failure of the licensee to develop a preventative maintenance schedule for inspections of safety-related equipment. Specifically, the licensee did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers (DCT). The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04930 and CR-WF3-2014-06100. The licensee developed preventative maintenance tasks to inspect the DCT tubes, including disassembly where necessary, to restore compliance.
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4Q/2014 Inspection Findings - Waterford 3 The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to inspect portions of the dry cooling towers prevented the licensee from identifying corrosion that eventually degraded the system enough to cause a leak. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. The inspectors concluded that the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee did not implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Design Control Measures for the Suitability of Safety-Related Relays The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not have an adequate replacement frequency for safety-related relays associated with engineered safety features equipment to ensure that all required equipment operated in the time sequence assumed by the safety analysis. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-05091. The licensee replaced the affected relays and reduced their replacement frequency from 18 years to 3 years to restore compliance.
The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop an adequate replacement frequency for the relays used to monitor for under-voltage conditions on the safety-related emergency busses could have prevented the equipment from performing its safety function. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system component and the affected equipment maintained its operability. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, Challenging the Unknown, because the licensee did not stop when faced with uncertain conditions and risks were not evaluated and managed before preceding.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Quality in a Timely Manner The inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a condition adverse to quality in a time commensurate with the safety significance of the issue. Specifically, the licensee failed to repair degraded conduit that had been identified as corroded since 2008. As a result, conduits that were housing cables for safety-related components were degraded to the point where water entered the conduit and submerged cables that Page 4 of 13
 
4Q/2014 Inspection Findings - Waterford 3 were not designed for submergence for an extended period of time. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04951. The licensee repaired the degraded conduit associated with the impacted safety-related cables to restore compliance, and also initiated an extent of condition review to identify other cables that could potentially be impacted by degraded conduits.
The inspectors determined that the performance deficiency was more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, safety-related cables that were not rated for full submergence were submerged in water since at least 2008, potentially affecting the integrity of the cable and potentially impacting the safety-related equipments ability to perform their safety function in the event of an accident. The inspectors determined that the finding had very low safety significance (Green) because the finding impacted the qualification of mitigating components but the components maintained operability. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee decision-making practices did not emphasize prudent choices over those that are simply allowable.
Specifically, when evaluating condition reports written through several years that document the degraded conduit, the licensee elected to defer needed maintenance instead of placing the adequate priority on the issue.
Inspection Report# : 2014005 (pdf)
Significance:      Jun 06, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedures for Securing Dry Cooling Tower Fans A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III occurred when the licensee did not assure that design basis information was translated into specifications, drawings, procedures, and instructions.
Specifically, after a failure revealed new design basis information regarding the need to place a train of dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet, the licensee failed to incorporate this information into procedures. As a result, the failure recurred. The licensee entered this condition into its corrective action program as Condition Reports CR-WF3-2012-05680 and -06908 and updated procedure OP-006-005, Inverters and Distribution, to incorporate the new design basis information into procedures. The licensee documented its failure to timely update design basis information in Condition Report CR-WF3-2014-02981.
The failure to assure that design basis information was translated into specifications, drawings, procedures, and instructions as required by 10 CFR Part 50, Appendix B, Criterion III was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate design basis information regarding the need to place the dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet into specifications, drawings, procedures, and instructions impacted the capability, availability, and reliability of both trains of dry cooling towers. Using NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because the required number of dry cooling towers in the protected train maintained their operability.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee had not taken effective corrective actions to address an issue in a timely manner commensurate with its safety significance (P.3).
Inspection Report# : 2014008 (pdf)
Significance:      Jun 06, 2014 Identified By: NRC Page 5 of 13
 
4Q/2014 Inspection Findings - Waterford 3 Item Type: NCV NonCited Violation Failure to Identify and Correct Condition Adversely Affecting Flooding Mitigation Design The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. On May 19, 2014, the team identified a significant amount of debris on the floor of one of the main steam isolation valve areas. In a probable maximum precipitation event, this debris could have prevented sufficient water removal by the floor drains to meet design basis assumptions. Following identification, the licensee entered this condition into its corrective action program as Condition Report CR-WF3-2014-03037 and removed the debris from the area.
Excessive debris in the main steam isolation valve A area that could challenge the waterremoval capability of safety-related drain systems was a condition adverse to quality. The licensees failure to promptly identify and correct this condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. The lead inspector performed the initial significance determination for performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because it involved the degradation of equipment specifically designed to mitigate a flooding event. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The bounding change to the core damage frequency was 4.7x10-8 per year (Green). The dominant core damage sequences included extremely heavy rainfall, a loss of offsite power initiating event, failure of the train B 4.16kV bus, and failure of the pressurizer safety relief valves to close. The low initiating event frequency reduced the risk significance.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensees corrective actions from the previous non-cited violation did not fully address the issue (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Correct Multiple Degraded or Nonconforming Conditions The inspectors identified multiple instances of the licensees failure to promptly correct degraded or nonconforming conditions as required by 10 CFR Part 50, Appendix B, Criterion XVI. At the conclusion of the inspection, the licensee had one structure, system or component that had been degraded since November 2008, requiring compensatory measures to provide reasonable assurance of operability; the licensee had another degraded condition that had existed since April 2011 with no compensatory measures in place. Following the teams identification of this issue, the licensee documented this issue in Condition Report CR-WF3-2014-03250 to evaluate the timeliness of its corrective actions.
The failure to promptly correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency is more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, the team determined this finding to be of very low safety significance (Green) because it did not represent the actual loss of function of a safety-related system or train.
This finding has an evaluation cross-cutting aspect in the problem identification and resolution cross-cutting area Page 6 of 13
 
4Q/2014 Inspection Findings - Waterford 3 because the licensee failed to thoroughly evaluate the issues to ensure that the resolutions addressed causes and extents of condition commensurate with the issues safety significance (P.2).
Inspection Report# : 2014008 (pdf)
Significance:        Feb 20, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Adequate Design Control Measures for the Selection and Review for the Suitability of Application of Molded Case Circuit Breakers A self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, occurred because the licensee did not establish design control measures for the selection and review for the suitability of application of a molded case circuit breaker that was essential to the safety-related function of a shutdown cooling heat exchanger fan cooler. Specifically, the licensee did not select and review for the suitability of the correct safety-related circuit breaker for the application to provide circuit fault protection to the train B shutdown cooling heat exchanger air handling unit fan motor. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-02316 and CR-WF3-2013-04644. The immediate corrective action taken to restore compliance included the replacement of the breaker with a breaker more suitable for the application to protect the air handling unit fan motor. The planned corrective actions included an extent of condition review for other installed breakers and the revision of work order instructions to eliminate the practice of substituting and using the factory acceptance testing for pre-installation and post-maintenance tests, respectively.
The inspectors concluded that the failure to establish design control measures for the selection and review for suitability of application for the correct safety-related circuit breaker was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect breaker affected the availability, reliability, and capability of the shutdown cooling heat exchanger fan coolers to remove heat from the shutdown cooling heat exchanger areas following a design basis accident. The inspectors performed the initial significance determination. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The total exposure period was 23 days. The allowed outage time was 7 days. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was 5E-13/year (Green). The dominant core damage sequences included loss of offsite power events, failure of both trains of containment spray, and the failure of a pressurizer safety relief valve to remain closed. The equipment that helped mitigate the risk included the emergency diesel generators and the essential feedwater systems.
The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect of avoiding complacency in the human performance area because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk on relying on 21 year old vendor information and installing a breaker without pre-installation and adequate post-maintenance testing.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 08, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Page 7 of 13
 
4Q/2014 Inspection Findings - Waterford 3 Failure to Replace an Essential Chiller Oil Pump prior to the End of Duty Life.
A Green self-revealing, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish preventative maintenance schedule to inspect or replace an item that have a specific lifetime. Specifically, the licensee did not establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the pump motor exceeding its duty life. As a result, the pump associated with essential chiller B failed in-service. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00095. The immediate corrective action taken to restore compliance was to issue an action request to establish the periodic replacement of the essential chiller pumps prior to the end of their vendor recommended service life.
The failure to establish a preventative maintenance schedule to inspect or replace the oil pump motors associated with the essential chillers prior to the end of the vendor provided duty life was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a preventative maintenance schedule to inspect or replace the oil pumps associated with the essential chillers prior to the duty life resulted in the failure of a pump while in service and the unavailability of essential chiller B. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 06, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Procedures for Using the Alternate Emergency Fuel Oil Storage Tank Fill Line.
An NRC-identified Green, non-cited violation of Technical Specification 6.8.1.a, occurred because the licensee did not establish written procedures for filling emergency power sources. Specifically, the licensee did not establish procedures to fill the fuel oil storage tanks for the emergency diesel generators using the safety related, seismic category I alternate emergency fill line. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-00636. The immediate corrective action taken to restore compliance was to initiate actions for developing procedures for filling the emergency diesel generator fuel oil storage tanks using the alternate emergency fill line.
The failure to develop procedures for filling emergency power sources was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the failure to establish procedures for the filling of the emergency diesel fuel oil storage tanks using the Seismic Category I alternate emergency fill connection reduced the licensees capability and reliability to for filling the fuel oil storage tanks following an extreme weather event. The inspectors inspector performed the initial significance determination and used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 2, Mitigating Systems Screening Questions, to evaluate this issue.
The finding required a detailed risk evaluation because the performance deficiency could have resulted in a loss of safety function (onsite AC power) because the system may not have remained operable for its 30 day design basis accident mission time. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation for this issue. The analyst determined that the finding was of very low safety significance (Green) because the diesel Page 8 of 13
 
4Q/2014 Inspection Findings - Waterford 3 generators would have remained functional for their 24-hour probabilistic risk assessment mission time. This shorter mission time is used for detailed risk evaluations because, after 24 hours, the NRC assumes that the licensee has substantially more resources available to help mitigate the accident. The dominant core damage sequences included longer term loss of offsite power events and the common cause failure of the diesel generators because of potential problems refilling the diesel fuel oil storage tanks. The relatively long period prior to ultimate diesel generator failure helped to minimize the risk. The finding was not a significant contributor to the large early release frequency. The inspectors concluded that the finding reflected current licensee performance and involved an avoiding complacency cross-cutting aspect of the human performance area in that the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014002 (pdf)
Significance:        Jan 03, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Evaluation for Transient Combustibles.
An NRC-identified Green, non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9 and the Fire Protection Program occurred because the licensee failed to follow procedures.
Specifically, the licensee did not perform a transient combustible evaluation as required by EN-DC-161, Control of Combustibles, to evaluate the impact of capturing and storing up to two gallons of leaking fuel oil in the train B emergency diesel generator room. As a result, the licensee was not performing required hourly fire watches. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-6020 and CR-WF3-2013-06123. The immediate corrective action taken to restore compliance was to perform a transient combustible evaluation implement hourly fire watches.
The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform a transient combustible evaluation when a flammable liquid above one pint in an approved container was present in the B emergency diesel generator room prevented the licensee from implementing required compensatory measures in response to the presence of transient combustibles. In addition, similar to NRC Inspection Manual Chapter 0612, Appendix E, Section 4, Example k of a more than minor violation, the failure of the leak collection device resulting in fuel oil around emergency diesel generator B represented a credible fire scenario involving transient combustibles that could affect equipment important to safety. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The inspectors categorized the finding under Fire Prevention and Administrative Controls and qualitatively screened it as very low safety significance (Green) because the impact of the fire finding was limited to no more than one train of equipment important to safety. The inspectors concluded that the finding reflected current licensee performance and involved a conservative bias cross-cutting aspect in the human performance area in that the licensee did not use decision making practices that emphasized prudent choices over those that are simply allowable.
Inspection Report# : 2014002 (pdf)
Significance:        Dec 20, 2013 Identified By: NRC Item Type: VIO Violation Failure to establish an adequate test program to demonstrate that the train B EDG exhaust fan would perform satisfactorily in service Page 9 of 13
 
4Q/2014 Inspection Findings - Waterford 3 A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criteria XI, Test Control, occurred because the licensee failed to establish an adequate test program to demonstrate that a safety related component associated with the Train B Emergency Diesel Generator would perform satisfactorily in service. Specifically, the licensee failed to identify and perform adequate testing on the Train B EDG exhaust fan to demonstrate that the exhaust fan would perform satisfactorily in service, which incorporated the requirements and acceptance limits contained in applicable design documents such as the Final Safety Analysis Report, as updated. As a result, the licensee failed to ensure that for all operational tests that the safety related exhaust fan would perform satisfactorily such that it would provide sufficient flow and remove heat during accident conditions. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-02530. The immediate corrective actions taken to restore compliance included the replacement of the B EDG exhaust fan assembly. The planned corrective actions include the review of the EDG ventilation system monitoring plan.
The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the licensee failed to perform testing to ensure that the B EDG ventilation exhaust fan would fulfill its safety function to remove heat from the EDG room when the diesel operates during accident conditions. The senior resident inspector performed the initial significance determination for the diesel generator room ventilation fan failure.
The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. The emergency diesel generator needed the ventilation exhaust fan to remain Operable. The unit was not recoverable. The total exposure period was 25 days. The allowed outage time was 72 hours. The analyst determined the best estimated change to the core damage frequency was 4.4E-6/year (White). The risk significance was low to moderate (White). The dominant core damage sequences included loss of offsite power events, leading to station blackout, coincident with the failure of the turbine driven auxiliary feedwater pump. Equipment that helped mitigated the risk included recovery of an emergency diesel generator or manually starting a temporary emergency diesel generator set. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the resource component of the human performance area in that the licensee did not have complete, accurate and up-to-date operational surveillance procedure tests [H.2.c].
Final significance determination and White NOV issued March 28, 2014. IR 05000382/2014009 (ML14086A768)
IP 95001 supplemental inspection performed. IR 05000382/2014011 (ML14364A412) issued December 30, 2014.
White finding is being held open due to IP 95001 objectives not being met due to inadequate extent of cause evaluation.
Inspection Report# : 2013008 (pdf)
Inspection Report# : 2014009 (pdf)
Inspection Report# : 2014011 (pdf)
Barrier Integrity Emergency Preparedness Page 10 of 13
 
4Q/2014 Inspection Findings - Waterford 3 Significance:        Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions.
The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860.
The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes.
Inspection Report# : 2014003 (pdf)
Occupational Radiation Safety Significance:        Dec 31, 2014 Identified By: NRC Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities Related to Alloy 600 Pipe Weld Inspections to Ensure Doses were ALARA.
The inspectors identified a finding of very low safety significance associated with the licensees failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations during Refueling Outage 19.
Specifically, the inspectors concluded that, had the licensee appropriately evaluated the Alloy 600 pipe weld conditions/ locations during the ALARA planning process and appropriately performed in-progress ALARA reviews, they could have reasonably planned for the full scope of work and provided a better estimate and/or adequately Page 11 of 13
 
4Q/2014 Inspection Findings - Waterford 3 justified revising the estimate for the job. These failures to plan and control the job activities led to unplanned, unintended collective dose. The licensee evaluated the procedures used during this work, including their process for planning and estimating doses, and documented the issue in the corrective action program.
The failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations is a performance deficiency. This performance deficiency is more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, it caused the collective radiation dose for the work to be greater than 5 man-rem and exceed the planned dose estimate by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was associated with ALARA planning and (2) the licensees three-year rolling average collective dose of 121.7 man-rem was less than 135 man-rem. The finding has a Work Management cross-cutting aspect, associated with the Human Performance cross-cutting area, because the licensee did not adequately plan or control work activities such that nuclear safety is the overriding safety priority. Specifically, the ALARA plan did not reflect the time needed to complete the work activities, thus underestimating the dose requirements, and the administrative control of reviewing the work-in-progress at appropriate completion points failed.
Inspection Report# : 2014005 (pdf)
Significance:        Jun 30, 2014 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control Entry into a High Radiation Area The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638.
The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014003 (pdf)
Public Radiation Safety Page 12 of 13
 
4Q/2014 Inspection Findings - Waterford 3 Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
Inspection Report# : 2014008 (pdf)
Last modified : February 26, 2015 Page 13 of 13
 
1Q/2015 Inspection Findings - Waterford 3 Waterford 3 1Q/2015 Plant Inspection Findings Initiating Events Significance:      Jun 06, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify a Cause and Implement Corrective Actions to Prevent Recurrence for a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for a failure to identify a cause and take corrective actions to prevent recurrence. Specifically, the licensee did not identify a cause or corrective actions to prevent recurrence for a plant trip and equipment failures caused by elevated main feed system vibrations.
The licensee replaced the steam generators at Waterford 3 during refueling outage 18 in late 2012. Upon returning to power operations the licensee experienced elevated vibration levels and related equipment failures on the main feedwater system and emergency feedwater system. The most significant of these failures included a plant trip after a loss of instrument air to the feedwater regulating valve actuator. The licensee determined that the plant trip was a significant event, and initiated a root cause evaluation through its corrective action process. This root cause determination identified a possible cause, which by the licensees program required additional information to confirm or refute. The licensee initiated a proposal to perform modeling of the steam generator flows to provide this information, but later canceled the action. No corrective actions to prevent recurrence were implemented by the licensee. Actions taken to date by the licensee appear to have been effective in mitigating known effects of the vibrations. The licensee documented its failure to determine and document the cause of these vibrations in Condition Report CR-WF3-2014-03238.
The failure to identify the cause of the feedwater vibration-induced problems and to take corrective actions to prevent recurrence as required by 10 CFR Part 50, Appendix B, Criterion XVI is a performance deficiency. The performance deficiency is more than minor because if left uncorrected, it could lead to a more significant safety concern.
Specifically, though individual actions were taken to address failures caused by vibrations, no actions were taken to reduce or eliminate the vibrations themselves. Actions that were taken were not treated as corrective actions to prevent recurrence. A lack of corrective actions to prevent recurrence could leave main feedwater components and other components physically connected to the system such as emergency feedwater susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the team determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the initiating events cornerstone, did not result in a reactor trip and the loss of mitigating equipment needed to transition the plant from the onset of the trip to a stable shutdown condition.
This finding has a resources cross-cutting aspect in the human performance area because leaders did not ensure that procedures used at the time the root cause assessment was performed were adequate to support nuclear safety (H.1).
The procedure used by the licensee allowed a root cause assessment to have an indeterminate root cause and thus no corrective actions to prevent recurrence.
Inspection Report# : 2014008 (pdf)
Significance:      Jun 06, 2014 Identified By: NRC Page 1 of 20
 
1Q/2015 Inspection Findings - Waterford 3 Item Type: FIN Finding Failure to Evaluate Operating Experience as Directed in Station Procedure The team identified a finding for the licensees failure to evaluate industry operating experience as directed in the station operating experience program procedure. Specifically, a vendor supplied Technical Bulletin TB-13-1 Steam Generator and Pressurizer Closure Gasket Replacement Frequency, which recommended that all Westinghouse-designed steam generator and pressurizer closure gaskets be replaced at a prescribed frequency, was not evaluated in accordance with station procedures. This resulted in the licensee failing to take action to periodically replace affected gaskets to preclude degradation of the pressure boundary. The licensee documented this performance deficiency in Condition Report CR-WF3-2014-03229 to determine what further actions were needed.
The failure to evaluate operating experience information as required by licensee procedure EN-OP-100, Operating Experience Program, Revision 20, was a performance deficiency. The performance deficiency is more than minor because if left uncorrected it would have the potential to lead to a more safety-significant concern. Specifically, the failure of the licensee to take any action with regard to the technical bulletin recommendation to replace the steam generator gaskets would allow the gaskets to be installed longer than their useful life. The deterioration of gasket material could result in unplanned transients or shutdowns. The finding is therefore associated with the initiating events cornerstone. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (Green) because it was not an actual degradation that could have resulted in exceeding a reactor system leak rate for a small LOCA; could not have affected other systems used to mitigate a LOCA; did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition; and did not involve a complete or partial loss of a support system that contributes to the likelihood of, or causes, an initiating event and affected mitigation equipment.
This finding has a conservative bias cross-cutting aspect in the human performance area (H.14). Specifically, the licensee assumed that the technical bulletin was not based on actual failures and because steam generators had just been replaced, opted not to take further actions to evaluate or initiate any preventative maintenance to replace gaskets.
Inspection Report# : 2014008 (pdf)
Mitigating Systems Significance:        Feb 21, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Perform Testing of Safety-Related Dry Cooling Tower Tube Bundle Isolation Valves The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not identify and perform testing for safety-related components to demonstrate that they would perform satisfactorily in service.
Specifically, prior to February 12, 2015, the licensee did not identify and perform testing to demonstrate that, as described in the licensees design basis, the dry cooling tower tube bundle isolation valves could be used to isolate a dry cooling tower tube bundle following a tornado missile strike on the non-missile-protected portions of the dry cooling tower.
The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2015-00828. The planned corrective actions are to develop seat leakage criteria for the dry cooling tower tube bundle isolation valves and to perform periodic seat leakage testing.
The inspectors determined that the performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Page 2 of 20
 
1Q/2015 Inspection Findings - Waterford 3 Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a test program for a safetyrelated component to demonstrate that it would perform satisfactorily following a tornado missile strike could impact the systems ability to perform its safety function in the event of a tornado. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Event Screening Questions. The finding required a detailed evaluation because it would degrade one or more trains of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 2.9E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado-induced losses of offsite power, failure of the train B dry cooling tower pressureboundary, random failure of the train A component cooling water system, random failures of the emergency diesel generators, and failure to recover offsite power in 4 hours. Risk was minimized because the diesel generators have air cooled radiators and do not require component cooling water to remain functional. The low tornado frequency also minimized the risk.
The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency of not identifying the need for a leak test occurred more than two years ago and did not reflect current licensee performance.
Inspection Report# : 2015001 (pdf)
Significance:        Feb 13, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Fire Area Boundary The team identified a non-cited violation of License Condition 2.C.(9), Fire Protection, for the failure to ensure the required separation between fire areas. Specifically, the licensee installed fire barriers on two ventilation ducts which were not in a configuration demonstrated to provide the required three-hour fire-rated separation between fire areas.
The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-00540 and established an hourly fire watch as a compensatory measure until corrective actions can be taken (Fire Impairments 15-30 and 15-31).
The failure to ensure the required separation between fire areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. Both emergency diesel generator rooms were equipped with pre-action sprinkler systems which would limit temperatures near the ceiling around the room exhaust ducts; therefore, the finding screened to Green at Section 1.4.3.C.
This finding did not have a cross-cutting aspect since it was not indicative of current licensee performance since this fire barrier configuration was installed in the 1980s.
Inspection Report# : 2015007 (pdf)
Page 3 of 20
 
1Q/2015 Inspection Findings - Waterford 3 Significance:        Feb 13, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Provide a Bounding Calculation for Time Critical Actions The team identified a non-cited violation of License Condition 2.C.9, "Fire Protection," for the failure to adequately correct a previous violation. Specifically, the licensee failed to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown. The licensee developed this calculation in response to Non-cited Violation 2012007-02. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-0859 and implemented a fire impairment as a compensatory measure.
The failure to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a postulated control room fire that led to control room evacuation and determined this violation was of very low safety significance.
This finding had a cross-cutting aspect associated with resolution within the problem identification and resolution area since the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the team determined that the licensees corrective actions were not effective since the licensee failed to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown (P.3).
Inspection Report# : 2015007 (pdf)
Significance:        Feb 13, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Periodically Test Emergency Lighting Units The team identified a non-cited violation of License Condition 2.C.9, Fire Protection, for the failure to periodically test and demonstrate the 8-hour capacity of the Appendix R emergency lighting units. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-00856 and operators had flashlights available as a compensatory measure.
The failure to periodically test and demonstrate the 8-hour capacity of the Appendix R emergency lighting units was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. The team assigned the finding a low degradation rating because it would not prevent reaching and maintaining safe shutdown conditions in the event of a control room fire. Specifically, the team had reasonable assurance that the emergency lighting units would provide adequate illumination for a sufficient amount of time for operators to perform the most time critical actions. In addition, the team determined that operators performing an alternative shutdown had flashlights available in the Appendix R equipment lockers. Because the team assigned a low degradation rating, this finding screened as having very low safety significance.
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1Q/2015 Inspection Findings - Waterford 3 This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.
Inspection Report# : 2015007 (pdf)
Significance:        Feb 13, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Correct Long Standing Deficiencies with the Appendix R Emergency Lighting Units The team identified a non-cited violation of License Condition 2.C.9, Fire Protection, for the failure to correct adverse conditions associated with fire protection. Specifically, the licensee failed to correct longstanding deficiencies with the Appendix R emergency lighting units. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-00593 and operators had flashlights available as a compensatory measure.
The failure to correct longstanding deficiencies with the Appendix R emergency lighting units was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. The team assigned the finding a low degradation rating because the failure to provide adequate 8-hour emergency lights at all locations would not prevent reaching and maintaining safe shutdown conditions in the event of a control room fire. Specifically, the team determined that operators performing an alternative shutdown had flashlights available in the Appendix R equipment lockers. Because the team assigned a low degradation rating, this finding screened as having very low safety significance.
This finding had a cross-cutting aspect associated with resolution within the problem identification and resolution area since the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the team determined that the licensee failed to take corrective actions to address the nonfunctional emergency lighting units in a timely manner (P.3).
Inspection Report# : 2015007 (pdf)
Significance:        Jan 12, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Evaluate Elevated Bus Voltages Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, during the periods of October 27 through December 13, 2012, and on May 1, 2014, the licensee failed to identify and evaluate the impact of elevated bus voltages that exceeded the allowable voltage on the 480 VAC Class 1E Bus 3B31, a condition adverse to quality. In response to this issue, the licensee completed an operability determination with plans to evaluate any trends requiring additional actions. This finding was entered into the licensees corrective action program as Condition Report CR WF3 2014-05458.
The team determined that the failure to identify and evaluate the impact of elevated bus voltages was a performance deficiency. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the Page 5 of 20
 
1Q/2015 Inspection Findings - Waterford 3 licensee failed to identify and evaluate elevated voltages on the 480 VAC Class 1E Bus 3B31 that exceeded allowable operability limits. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution associated with trending because the licensee failed to periodically analyze information in the aggregate to identify programmatic and common cause issues. [P.4](Section 1R21.2.2)
Inspection Report# : 2014007 (pdf)
Significance:        Jan 12, 2015 Identified By: NRC Item Type: FIN Finding Inadequate Station Procedures for Temporary Emergency Diesel Generator Green. The team identified a Green finding for inadequate station procedures for the temporary emergency diesel generators. Specifically, the licensee failed to ensure that Procedures OP-TEM-008, Emergency Diesel Generator A (B) Backup Temporary Diesel Generators, and ME-001-012, Temporary Power from Temporary Diesel for 3A2 and 3B2 4kV Buses (MODES 1-6), were maintained to ensure that the temporary diesels had enough capacity to supply auxiliary power to the required safe-shutdown loads. The team determined that the licensee failed to clearly establish appropriate instructions to ensure that operators would be running and verifying loads according to the prime rating, that three temporary diesels were capable of operating/connecting in parallel, and that required and desired loads were consistent between procedures and evaluations. In response to this issue, the licensee evaluated and updated station procedures, specified prime loading limitations, updated vendor contract, incorporated procedure improvements as a result of training, and updated the adverse weather procedure. This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2014-05662 and CR WF3 2014 05582.
The team determined that failure to maintain procedures that ensure the temporary diesels have enough capacity to supply auxiliary power to required safe-shutdown loads was included in station procedures was a performance deficiency. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to update Procedures OP TEM 008 and ME-001-012, and vendor documents in accordance with engineering evaluation EC-47496, in a timely manner and prior to performance of the emergency diesel generator outage in January 2014. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance associated with teamwork because the licensee failed to ensure that individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. [H.4](Section 1R21.2.7)
Inspection Report# : 2014007 (pdf)
Significance:        Jan 12, 2015 Page 6 of 20
 
1Q/2015 Inspection Findings - Waterford 3 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Initiate a Condition Report for a Condition Adverse to Quality Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, between October 8 and 16, 2014, the licensee failed to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents, a nonconformance that is a condition adverse to quality for eight days. In response to this issue, the licensee performed an operability determination to address the teams concerns and initiated a separate condition report to document the lack of initiating and evaluating a condition report for a condition adverse to quality in a timely manner. This finding was entered into the licensees corrective action program as Condition Reports CR WF3 2014-05341 and CR WF3 2014 05738.
The team determined that the failure to initiate a condition report to evaluate the lack of missile protection on the emergency diesel generator A and B storage tank vents for eight days was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to initiate and evaluate a condition adverse to quality, a design nonconformance on the emergency diesel generator A and B storage tank vents for missile protection for eight days. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather.
This finding had a crosscutting aspect in the area of human performance associated with work management because the licensee failed to implement a process where nuclear safety is the overriding priority and the need for coordinating with different work groups. [H.5](Section 1R21.2.12.1)
Inspection Report# : 2014007 (pdf)
Significance:        Jan 12, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Evaluate Missile Protection Requirements for Emergency Diesel Generator Day and Storage Tank Vents Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to November 6, 2014, the licensee did not verify the adequacy of design of the emergency diesel generator A and B day and storage tank vents to have missile protection installed, or an approved exemption excluding missile protection requirements. In response to this issue, the licensee performed a TORMIS evaluation that supported an operable determination and a future licensing basis change. TORMIS is an EPRI methodology documented in EPRI NP 2005, Tornado Missile Simulation and Design Methodology, dated August 1981, and was approved for use by Waterford in the Safety Evaluation related to License Amendment 168. This finding was entered into the licensees corrective action program as Condition Reports CR WF 2014 05131, CR WF3 2014 5341, and CR-WF3-2014-5412.
The team determined that the failure to evaluate the lack of missile protection on the emergency diesel generator A Page 7 of 20
 
1Q/2015 Inspection Findings - Waterford 3 and B day and storage tank vents was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to evaluate a design nonconformance on the emergency diesel generator A and B day and storage tank vents for lack of missile protection. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.12.2)
Inspection Report# : 2014007 (pdf)
Significance:      Jan 12, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Correct Through Wall Corrosion on Emergency Diesel Generator A and B Day Tank Vents
* TBD. The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, prior to October 22, 2014, the licensee failed to identify and correct through wall corrosion on the emergency diesel generator A and B day tank vents, a condition adverse to quality. The team asked the licensee if the corrosion had been previously evaluated. The licensee determined that it had not been aware of the corrosion so it had not been evaluated. The corrosion was significant enough that a through wall hole had formed at the base of the each vent pipe where it penetrates the roof. Consequently, any water that collects on the roof of the building would have the potential to drain into the respective day tank. In response to this issue, the licensee performed an immediate operability determination based on severe weather in the area, installed a temporary repair using a rubber wrap, and installed a small concrete berm to minimize the potential amount of water in the immediate area. This finding was entered in to the licensees corrective action program as Condition Report CR WF3 2014 05413.
The team determined that the failure to identify and correct through wall corrosion on the emergency diesel generator A and B day tank vents was a performance deficiency. This finding was more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to identify, evaluate, and correct through wall corrosion on the emergency diesel generator A and B day tank vents. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened to Exhibit 4, External Events Screening Questions, because it screened as potentially risk significant due to seismic, flooding, or severe weather.
Per Exhibit 4 the issue screened to a Detailed Risk Evaluation because if the safety function were assumed completely failed, emergency diesel generator A and B, it would degrade two trains of a multi-train system and it would degrade one or more trains of a system that supports a risk significant system.
A Region IV senior reactor analyst performed a detailed risk evaluation. The finding was potentially Greater than Green in significance and the NRC requested the licensee to provide additional information to enable the NRC to determine the final significance. The risk important sequences included heavy rain induced losses of offsite power Page 8 of 20
 
1Q/2015 Inspection Findings - Waterford 3 with the consequential failure of both emergency diesel generators. The ability to restore offsite power within 4 hours was important to avoid core damage. The finding was not significant to the large early release frequency. See , Detailed Risk Evaluation, for a detailed review of the Appendix M evaluation.
This finding had a crosscutting aspect in the area of human performance associated with procedure adherence because the licensee failed to ensure that individuals follow process, procedures, and work instructions. [H.8](Section 1R21.2.12.3)
(Update)
The finding was determined to be of very low safety significance (Green), in part based on the licensees testing of the roof drain and the Cooper Bessemer diesel tolerance to water. The change to the core damage frequency was approximately 4x10-7/year. The risk-important sequences included a heavy rain event greater than or equal to 6 inches per hour followed by a random loss of offsite power within the next two weeks. The risk significance was mitigated by the tolerance of the diesel generators to water in the fuel oil and the operators ability to restore offsite power within 4 hours of the loss of offsite power. (IR 05000382/2015001 and 05000382/2015009 dated May 14, 2015)
Inspection Report# : 2014007 (pdf)
Inspection Report# : 2015009 (pdf)
Significance:      Jan 09, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Procedure for Tightening Thermal Overload Connections for Safety-Related Components A self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, was identified for the failure to perform maintenance that could affect the performance of safety-related equipment in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Specifically, prior to December 17, 2014, the licensee used a procedure that contained insufficient detail for tightening a thermal overload connection that resulted in a loose connection on a motor starter and eventual trip of a wet cooling tower fan, resulting in the A train of ultimate heat sink being declared inoperable. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2014-04430.
The corrective action taken to restore compliance was to add additional detail to the procedure to ensure thermal overload connections are verified secure after their mechanical connections are tightened.
The inspectors determined that the performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure successful tightening of the thermal overload connections for the wet cooling tower fans adversely impacted the capability of the system to perform its function. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. The inspectors determined the finding was of very low safetysignificance (Green) because it affected one train for less than the allowed outage time.
When the A train of ultimate heat sink was declared inoperable, the B train of ultimate heat sink was already inoperable for planned maintenance. As a result, the B train maintenance was unrelated to the performance deficiency. In addition, the finding did not affect the Page 9 of 20
 
1Q/2015 Inspection Findings - Waterford 3 design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment.
The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency occurred more than two years agoand did not reflect current licensee performance.
Inspection Report# : 2015001 (pdf)
Significance:      Jan 08, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Correct a Condition Adverse to Fire Protection The inspectors identified a finding of very low safety significance and an associated non-cited violation of Waterford Steam Electric Station, Unit 3, License Condition 2.C.9, and the fire protection program for the licensees failure to identify and correct a condition adverse to fire protection. Specifically, the inspectors identified that the ventilation dampers that are used to maintain the environmental conditions of the No. 2 diesel fire pump room and that are needed for pump protection were damaged and not functional for an extended period of time. As a result, the reliability of the No. 2 diesel fire pump could have been impacted at high environmental temperatures. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2015-00132. The licensee manually opened the dampers and additional planned corrective actions included repairing the broken dampers linkage before the temperatures outside reach 90ºF.
This performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the licensees failure to repair the damaged ventilation damper in the No. 2 diesel fire pump room would result in an ongoing degraded condition, which could have impacted the capability of the No. 2 diesel fire pump to fulfill its function of providing a water supply to the sites Fire Protection Systems. Using Inspectional Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the use of Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was required because the finding involved fixed fire protection systems. Using Inspection Manual Chapter 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the finding screened as Green because the reactor would have been able to reach and maintain a safe shutdown condition. Specifically, since only the No. 2 diesel fire pump was impacted by the performance deficiency, the No. 1 diesel fire pump and the motor driven pump would have been able to supply the fire systems because they are all rated for full flow capacity.
This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, licensee personnel frequently tour the fire pump house for operations and maintenance activities; however, a thorough review of the work site had not been performed.
Inspection Report# : 2015001 (pdf)
Significance:      Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Page 10 of 20
 
1Q/2015 Inspection Findings - Waterford 3 Failure to Identify and Control Potential Tornado-Borne Missile Hazards The inspectors identified a non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 312, on two separate instances. Specifically, on both November 16 and December 23, 2014, the licensee entered the off-normal procedure due to a tornado watch but failed to assess and control potential tornado-borne missile hazards on site as required by the procedure. The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-05912 and CR-WF3-2014-06453. The immediate corrective action taken to restore compliance was to secure the identified hazards.
This finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, in the event of a tornado at the site, these loose items could have become missiles with the potential to impact safety-related site equipment and personnel. The inspectors determined the finding was of very low safety significance (Green) because the it did not involve the loss or degradation of equipment or functions specifically designed to mitigate a seismic, flooding, or severe weather event (e.g. seismic snubbers, flooding barriers, tornado doors). The inspectors concluded that the finding had a cross-cutting aspect in the area of Human Performance, Field Presence, because the licensee did not ensure supervisory and management oversight of work activities.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Follow the Operability Determination Process TThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to assess immediate operability of safety-related systems in accordance with site procedures, in three separate instances. Specifically, on two occasions, the licensee did not properly assess operability of safety-related relays in the Engineered Safety Features Actuation Signal system, which in turn brought into question the operability of the emergency diesel generators. A third example involved the licensees failure to properly assess operability of safety-related class 3 piping on the dry cooling towers, which brought into question the operability of the component cooling water system. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-06014. The licensee restored compliance by revising the immediate operability determinations to reflect an adequate reason to justify operability of the systems in questions.
The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to follow the Operability Determination procedure caused the licensee to incorrectly assess the capability of the systems impacted by the relays and dry cooling tower tube leak to perform their safety function and there was a reasonable doubt on the operability of the systems. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process, because individuals did not use a consistent, systematic approach to make a decision and risk insights were not incorporated appropriately.
Inspection Report# : 2014005 (pdf)
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1Q/2015 Inspection Findings - Waterford 3 Significance:      Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Establish an Inspection Schedule of the Dry Cooling Towers The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for failure of the licensee to develop a preventative maintenance schedule for inspections of safety-related equipment. Specifically, the licensee did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers (DCT). The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04930 and CR-WF3-2014-06100. The licensee developed preventative maintenance tasks to inspect the DCT tubes, including disassembly where necessary, to restore compliance.
The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to inspect portions of the dry cooling towers prevented the licensee from identifying corrosion that eventually degraded the system enough to cause a leak. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. The inspectors concluded that the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee did not implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Establish Design Control Measures for the Suitability of Safety-Related Relays The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not have an adequate replacement frequency for safety-related relays associated with engineered safety features equipment to ensure that all required equipment operated in the time sequence assumed by the safety analysis. The licensee entered this condition into their corrective action program as condition report CR-WF3-2013-05091. The licensee replaced the affected relays and reduced their replacement frequency from 18 years to 3 years to restore compliance.
The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop an adequate replacement frequency for the relays used to monitor for under-voltage conditions on the safety-related emergency busses could have prevented the equipment from performing its safety function. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system component and the affected equipment maintained its operability. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, Challenging the Unknown, because the licensee did not stop when faced with uncertain conditions and risks were not evaluated and managed before preceding.
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1Q/2015 Inspection Findings - Waterford 3 Inspection Report# : 2014005 (pdf)
Significance:      Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Correct a Condition Adverse to Quality in a Timely Manner The inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a condition adverse to quality in a time commensurate with the safety significance of the issue. Specifically, the licensee failed to repair degraded conduit that had been identified as corroded since 2008. As a result, conduits that were housing cables for safety-related components were degraded to the point where water entered the conduit and submerged cables that were not designed for submergence for an extended period of time. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04951. The licensee repaired the degraded conduit associated with the impacted safety-related cables to restore compliance, and also initiated an extent of condition review to identify other cables that could potentially be impacted by degraded conduits.
The inspectors determined that the performance deficiency was more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, safety-related cables that were not rated for full submergence were submerged in water since at least 2008, potentially affecting the integrity of the cable and potentially impacting the safety-related equipments ability to perform their safety function in the event of an accident. The inspectors determined that the finding had very low safety significance (Green) because the finding impacted the qualification of mitigating components but the components maintained operability. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee decision-making practices did not emphasize prudent choices over those that are simply allowable.
Specifically, when evaluating condition reports written through several years that document the degraded conduit, the licensee elected to defer needed maintenance instead of placing the adequate priority on the issue.
Inspection Report# : 2014005 (pdf)
Significance:      Jun 06, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Inadequate Procedures for Securing Dry Cooling Tower Fans A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III occurred when the licensee did not assure that design basis information was translated into specifications, drawings, procedures, and instructions.
Specifically, after a failure revealed new design basis information regarding the need to place a train of dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet, the licensee failed to incorporate this information into procedures. As a result, the failure recurred. The licensee entered this condition into its corrective action program as Condition Reports CR-WF3-2012-05680 and -06908 and updated procedure OP-006-005, Inverters and Distribution, to incorporate the new design basis information into procedures. The licensee documented its failure to timely update design basis information in Condition Report CR-WF3-2014-02981.
The failure to assure that design basis information was translated into specifications, drawings, procedures, and instructions as required by 10 CFR Part 50, Appendix B, Criterion III was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate design basis information regarding the need to place the dry cooling tower fan controllers to the off Page 13 of 20
 
1Q/2015 Inspection Findings - Waterford 3 position prior to de-energizing the associated control cabinet into specifications, drawings, procedures, and instructions impacted the capability, availability, and reliability of both trains of dry cooling towers. Using NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because the required number of dry cooling towers in the protected train maintained their operability.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee had not taken effective corrective actions to address an issue in a timely manner commensurate with its safety significance (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Correct Condition Adversely Affecting Flooding Mitigation Design The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. On May 19, 2014, the team identified a significant amount of debris on the floor of one of the main steam isolation valve areas. In a probable maximum precipitation event, this debris could have prevented sufficient water removal by the floor drains to meet design basis assumptions. Following identification, the licensee entered this condition into its corrective action program as Condition Report CR-WF3-2014-03037 and removed the debris from the area.
Excessive debris in the main steam isolation valve A area that could challenge the waterremoval capability of safety-related drain systems was a condition adverse to quality. The licensees failure to promptly identify and correct this condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. The lead inspector performed the initial significance determination for performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because it involved the degradation of equipment specifically designed to mitigate a flooding event. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The bounding change to the core damage frequency was 4.7x10-8 per year (Green). The dominant core damage sequences included extremely heavy rainfall, a loss of offsite power initiating event, failure of the train B 4.16kV bus, and failure of the pressurizer safety relief valves to close. The low initiating event frequency reduced the risk significance.
This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensees corrective actions from the previous non-cited violation did not fully address the issue (P.3).
Inspection Report# : 2014008 (pdf)
Significance:        Jun 06, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Promptly Correct Multiple Degraded or Nonconforming Conditions The inspectors identified multiple instances of the licensees failure to promptly correct degraded or nonconforming conditions as required by 10 CFR Part 50, Appendix B, Criterion XVI. At the conclusion of the inspection, the licensee had one structure, system or component that had been degraded since November 2008, requiring compensatory measures to provide reasonable assurance of operability; the licensee had another degraded condition Page 14 of 20
 
1Q/2015 Inspection Findings - Waterford 3 that had existed since April 2011 with no compensatory measures in place. Following the teams identification of this issue, the licensee documented this issue in Condition Report CR-WF3-2014-03250 to evaluate the timeliness of its corrective actions.
The failure to promptly correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency is more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, the team determined this finding to be of very low safety significance (Green) because it did not represent the actual loss of function of a safety-related system or train.
This finding has an evaluation cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate the issues to ensure that the resolutions addressed causes and extents of condition commensurate with the issues safety significance (P.2).
Inspection Report# : 2014008 (pdf)
Barrier Integrity Significance:        Jan 12, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Properly Evaluate Main Feedwater Isolation Valve Required Thrust Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, since January 18, 2006, the licensee failed to evaluate the adequacy of design for the required thrust for the main feedwater isolation valves in accordance with the licensees analysis methodology presented in EPRI TR 103237-R2, EPRI MOV Performance Prediction Program. In response to this issue, the licensee recalculated the required thrust and performed an evaluation of the remaining margin on the main feedwater isolation valves that supported an operable determination. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2014-05690.
The team determined that the failure to evaluate the required thrust for the main feedwater isolation valves, assuming an appropriate valve disk to seat coefficient of friction, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the incorrect coefficient of friction assumption resulted in a reasonable question of operability of the main feedwater isolation valves to operate under design basis conditions; during a main steam line break when auxiliary feedwater was supplying inventory to the steam generators. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.15)
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1Q/2015 Inspection Findings - Waterford 3 Inspection Report# : 2014007 (pdf)
Significance:        Jan 12, 2015 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Properly Evaluate Main Steam Isolation Valve Weak Link The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Specifically, since January 18, 2006, the licensee has failed to evaluate the adequacy of design of the main feedwater isolation valve operators to provide adequate thrust in accordance with the licensees analysis methodology described in EPRI topical report TR 103237-R2, EPRI MOV Performance Prediction Program. In response to this issue, the licensee recalculated the required thrust and performed an evaluation that supported a determination that the valves remained operable. This finding was entered into the licensees corrective action program as CR WF3-2014-05690.
The team determined that the failure to evaluate the required thrust for operation of the main feedwater isolation valves, assuming an appropriate valve-disk-to-seat coefficient of friction, was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events.
Specifically, the incorrect coefficient of friction assumption resulted in a reasonable question of operability of the main feedwater isolation valves to operate under the design basis condition of a main steam line break while auxiliary feedwater is supplying inventory to the steam generators. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.15)
* Green. The team reviewed a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to the failure of main steam isolation valve MS-124A on January 5, 2013, the licensee failed to have an adequate weak-link evaluation for the main steam isolation valves. In response to this event, the licensee performed a seismic weak-link evaluation of the main steam isolation valves that supported a determination that the valves were operable. This finding was entered into the licensees corrective action program as CR-WF3-2014-05708.
The team determined that the failure to evaluate the main steam isolation valve maximum allowed thrust, assuming appropriate values for the structural limitations of the valve and actuator, was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events.
Specifically, the licensee used a non-conservative value for the maximum allowed thrust, and the error resulted in a failure of main steam isolation valve MS-124A, because the allowable nitrogen pressure for the valve actuator was Page 16 of 20
 
1Q/2015 Inspection Findings - Waterford 3 inappropriate. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, this finding screened as having very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2014007 (pdf)
Emergency Preparedness Significance:        Jun 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions.
The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860.
The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the 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. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes.
Inspection Report# : 2014003 (pdf)
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1Q/2015 Inspection Findings - Waterford 3 Occupational Radiation Safety Significance:      Dec 31, 2014 Identified By: NRC Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities Related to Alloy 600 Pipe Weld Inspections to Ensure Doses were ALARA.
The inspectors identified a finding associated with the licensees failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations during Refueling Outage 19. Specifically, the inspectors concluded that, had the licensee appropriately evaluated the Alloy 600 pipe weld conditions/locations during the ALARA planning process and appropriately performed in-progress ALARA reviews, they could have reasonably planned for the full scope of work and provided a better estimate and/or adequately justified revising the estimate for the job. These failures to plan and control the job activities led to unplanned, unintended collective dose. The licensee evaluated the procedures used during this work, including their process for planning and estimating doses, and documented the issue in the corrective action program.
The failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations is a performance deficiency. This performance deficiency is more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, it caused the collective radiation dose for the work to be greater than 5 man-rem and exceed the planned dose estimate by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding has very low safety significance because: (1) it was associated with ALARA planning and (2) the licensees three-year rolling average collective dose of 121.7 man-rem was less than 135 man-rem. The finding has a Work Management cross-cutting aspect, associated with the Human Performance cross-cutting area, because the licensee did not adequately plan or control work activities such that nuclear safety is the overriding safety priority.
Specifically, the ALARA plan did not reflect the time needed to complete the work activities, thus underestimating the dose requirements, and the administrative control of reviewing the work-in-progress at appropriate completion points failed.
Inspection Report# : 2014005 (pdf)
Significance:      Jun 30, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Control Entry into a High Radiation Area The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological Page 18 of 20
 
1Q/2015 Inspection Findings - Waterford 3 controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638.
The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Inspection Report# : 2014003 (pdf)
Public Radiation Safety Significance:        Jan 14, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Develop the Transportation Security Plan The inspectors identified a non-cited violation of 10 CFR 71.5, Transportation of Licensed Material, and 49 CFR 172, Subpart I, Safety and Security Plans. Specifically, licensee personnel failed to adequately develop their transportation security plan. This resulted in three Category 2 shipments being transported on public highways without security risk assessments being performed. The planned corrective actions were still being evaluated. The inspectors determined that no immediate safety concern existed because the shipments that had been made were received with no issues and the licensee had no pending Category 2 or higher shipments. The licensee documented the issue in its corrective action program as Condition Report CR-W3-2015-00506.
The licensees failure to adequately develop their transportation security plan is a performance deficiency. Procedure EN-RW-106, Integrated Transportation Security Plan, did not include all the components required by 49 CFR 172.802, Components of a Security Plan. The performance deficiency is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of public health and safetyfrom exposure to radioactive materials released into the public domain. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined the finding has very low safety significance (Green) because Waterford had an issue involving transportation of radioactive waste, but it did not involve: (1) a radiation limit being exceeded, (2) a breach of package during transport, (3) a certificate of compliance issue, (4) a low level burial ground nonconformance, or (5) a failure to make notifications or provide emergency information.
The finding has a resources cross-cutting aspect in the human performance cross-cutting area, because licensee management did not ensure that personnel, equipment, procedures, Page 19 of 20
 
1Q/2015 Inspection Findings - Waterford 3 and other resources were available and adequate to support nuclear safety.
Inspection Report# : 2015001 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Sep 30, 2013 Identified By: NRC Item Type: VIO Violation Failure to Make a Report Required by 10 CFR 21.21 The team identified a violation of 10 CFR 21.21 that occurred when the licensee failed to submit a report or interim report on a deviation in a basic component within 60 days of discovery.
The failure of the licensee to adequately evaluate deviations in basic components and to report defects is a performance deficiency. The NRCs significance determination process (SDP) considers the safety significance of findings by evaluating their potential safety consequences. This performance deficiency was of minor safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. Supplement VII to the version of the NRC Enforcement Policy that was in effect at the time the violation was identified provided as an example of a violation of significant regulatory concern (Severity Level III), An inadequate review or failure to review such that, if an appropriate review had been made as required, a 10 CFR Part 21 report would have been made. Based on this example, the NRC determined that the violation met the criteria to be cited as a Severity Level III violation. However, because of the circumstances surrounding the violation, including the removal from service of the affected components by an unrelated manufacturers recall, the severity of the cited violation is being reduced to Severity Level IV. Cross-cutting aspects are not assigned to traditional enforcement violations.
Inspection Report# : 2013004 (pdf)
Inspection Report# : 2014008 (pdf)
Last modified : June 16, 2015 Page 20 of 20
 
2Q/2015 Inspection Findings - Waterford 3 Waterford 3 2Q/2015 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Feb 21, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify and Perform Testing of Safety-Related Dry Cooling Tower Tube Bundle Isolation Valves The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not identify and perform testing for safety-related components to demonstrate that they would perform satisfactorily in service.
Specifically, prior to February 12, 2015, the licensee did not identify and perform testing to demonstrate that, as described in the licensees design basis, the dry cooling tower tube bundle isolation valves could be used to isolate a dry cooling tower tube bundle following a tornado missile strike on the non-missile-protected portions of the dry cooling tower.
The licensee entered this condition into their correctiv}}

Latest revision as of 13:54, 29 November 2024

2017 Q1-Q4 ROP Inspection Findings
ML20311A319
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Issue date: 11/07/2019
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