IR 05000315/2002009: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
 
(One intermediate revision by the same user not shown)
Line 1: Line 1:
{{Adams
{{Adams
| number = ML030280726
| number = ML030630823
| issue date = 01/27/2003
| issue date = 01/10/2003
| title = IR 05000315-02-009 (DRP) & IR 05000316-02-009 (Drp), on 10/01/2002-12/28/2002, Indiana Michigan Power Company. Non-Cited Violations Noted
| title = Memo 1/10/03, D.C. Cook, DRS Input to DRP IR 05000315-02-009 and 05000316-02-009
| author name = Collins L
| author name = Slawinski W
| author affiliation = NRC/RGN-III/DRP/RPB6
| author affiliation = NRC/RGN-III/DRS/PSB
| addressee name = Bakken A
| addressee name = Vegel A
| addressee affiliation = American Electric Power Co
| addressee affiliation = NRC/RGN-III/DRP/RPB6
| docket = 05000315, 05000316
| docket = 05000315, 05000316
| license number = DPR-058, DPR-074
| license number = DPR-058, DPR-074
| contact person =  
| contact person =  
| document report number = IR-02-009
| case reference number = FOIA/PA-2003-0379, FOIA/PA-2003-0388
| document type = Inspection Report, Letter
| document report number = IR-01-009
| page count = 78
| document type = Memoranda
| page count = 10
}}
}}


Line 18: Line 19:


=Text=
=Text=
{{#Wiki_filter:ary 27, 2003
{{#Wiki_filter:January 10, 2003 MEMORANDUM TO:
Anton Vegel, Chief Reactor Projects Branch 6 Division of Reactor Projects FROM:
Wayne Slawinski, Acting Chief /RA/
Plant Support Branch Division of Reactor Safety SUBJECT:
D. C. COOK NUCLEAR POWER PLANT DRS INPUT TO INTEGRATED REPORT 50-315/02-09; 50-316/02-09 Attached is the report input for D. C. Cook Nuclear Power Plant, Inspection Report 50-315/02-09; 50-316/02-0 This input provides the results of a recent inspection into aspects of the licensees Occupational Radiation Safety Progra The inspection focused on the licensees Radiological Access Control and Radiation Monitoring Instrumentation and Protective Equipment programs, both under the occupational radiation safety cornerston Also, the performance indicator for the Barrier Integrity cornerstone (i.e., Reactor Coolant System, Specific Activity) was verifie I have reviewed this input and have determined it is ready for distribution to the licensee and dissemination to the public.


==SUBJECT:==
After your concurrence, please return this hard copy to me and we will email the electronic version to you.
D.C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2 NRC INSPECTION REPORT 50-315/02-09(DRP); 50-316/02-09(DRP)


==Dear Mr. Bakken:==
Attachment:
On December 28, 2002, the NRC completed an inspection at your D. C. Cook Nuclear Power Plant, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on January 3, 2003, with Mr. J. Pollock and other members of your staff.
Input to Inspection Report 50-315/02-09; 50-316/02-09 CONTACT:
Ronald V. Schmitt, DRS (630) 829-9718 DOCUMENT NAME: G:\\DRS\\ML030630823.wpd To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RIII RIII RIII NAME RSchmitt:aa WSlawinski AVegel DATE 1/9/03 1/9/03 1/10/03 OFFICIAL RECORD COPY


This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
Cover Letter

No input, no significant finding Input below, no color or green findings were identified.


The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Title page Inspector:
 
Ronald Schmitt, Radiation Specialist
Based on the results of this inspection, six findings of very low safety significance (Green) were identified which involved violations of NRC requirements. However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations, in accordance with Section VI.A.1 of the NRC Enforcement Policy. If you contest the Non-Cited Violations, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the D. C. Cook facility.
 
Since the terrorist attacks on September 11, 2001, the NRC has issued two Orders (dated February 25, 2002, and January 7, 2003) and several threat advisories to licensees of commercial power reactors to strengthen licensee capabilities, improve security force readiness, and enhance access authorization. The NRC also issued Temporary Instruction 2515/148 on August 28, 2002, that provided guidance to inspectors to audit and inspect licensee implementation of the interim compensatory measures (ICMs) required by the February 25th Order. Phase 1 of TI 2515/148 was completed at all commercial nuclear power plants during calendar year (CY) 02, and the remaining inspections are scheduled for completion in CY 03. Additionally, table-top security drills were conducted at several licensees to evaluate the impact of expanded adversary characteristics and the ICMs on licensee protection and mitigative strategies. Information gained and discrepancies identified during the audits and drills were reviewed and dispositioned by the Office of Nuclear Security and Incident Response. For CY 03, the NRC will continue to monitor overall safeguards and security controls, conduct inspections, and resume force-on-force exercises at selected power plants.
 
Should threat conditions change, the NRC may issue additional Orders, advisories, and temporary instructions to ensure adequate safety is being maintained at all commercial power reactors.
 
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
 
Sincerely,
/RA/
Laura Collins, Acting Chief Branch 6 Division of Reactor Projects Docket Nos. 50-315; 50-316 License Nos. DPR-58; DPR-74
 
===Enclosure:===
Inspection Report 50-315/02-09(DRP);
50-316/02-09(DRP)
 
REGION III==
Docket Nos: 50-315; 50-316 License Nos: DPR-58; DPR-74 Report No: 50-315/02-09(DRP); 50-316/02-09(DRP)
Licensee: Indiana Michigan Power Company Facility: D. C. Cook Nuclear Power Plant, Units 1 and 2 Location: 1 Cook Place Bridgman, MI 49106 Dates: October 1, 2002 through December 28, 2002 Inspectors: B. Kemker, Senior Resident Inspector I. Netzel, Resident Inspector R. Azua, Project Engineer, Region IV M. Bielby, Operations Engineer J. Ellegood, Resident Inspector, Perry R. Gibbs, Senior Reactor Analyst, NRR R. Jickling, Emergency Preparedness Inspector R. Krsek, Resident Inspector, Palisades J. Maynen, Physical Security Inspector W. Poertner, Operations Engineer, NRR R. Powell, Senior Resident Inspector, Perry P. Prescott, Operations Engineer, NRR R. Schmidt, Radiation Specialist H. Walker, Senior Reactor Engineer R. Winter, Reactor Engineer S. Wong, Senior Reactor Analyst, NRR D. Wrona, Operations Engineer, NRR Approved by: L. Collins, Acting Chief Branch 6 Division of Reactor Projects


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
IR 05000315-02-09(DRP), IR 05000316-02-09(DRP), on 10/01/2002-12/28/2002, Indiana
ADAMS boilerplate - Inspectable Area: Occupational Radiation Safety


Michigan Power Company, D. C. Cook Nuclear Power Plant, Units 1 and 2. Maintenance Effectiveness, Identification and Resolution of Problems, Event Follow-up.
Modify second paragraph as follows:
 
The baseline inspection was conducted by a regional radiation specialist inspector.
This report covers a 13-week period of inspection by resident, regional, and headquarters based inspectors. The inspectors identified six Green findings. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process," (SDP). The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG 1649, "Reactor Oversight Process," Revision 3, dated July 2000.
 
A.      Inspector Identified Findings
 
===Cornerstone: Initiating Events===
 
C
: '''Green.'''
The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to assure that prompt corrective actions were taken to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures, a significant condition adverse to quality. This issue was self-revealed on May 12, 2002, when an automatic reactor trip of Unit 2 occurred due to the failure of redundant 24-volt direct current power supplies in reactor control instrumentation cabinet 2-PS-CGC-16. The failure of both power supplies caused the number 21 steam generator feedwater regulating valve to close.
 
Unit 2 subsequently tripped on low steam generator water level coincident with low feedwater flow.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this issue, if left uncorrected, would become a more significant safety concern with the likelihood of continued failures of reactor control instrumentation power supplies and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective. Specifically, the failure of redundant power supplies in reactor control instrumentation cabinets would upset plant stability (cause a reactor trip) and challenge the function of critical safety equipment. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A,
                "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding contributes to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available, the inspectors determined that this finding required a Phase 2 SDP analysis.
 
After a review of additional information, the inspectors determined that a Phase 3 analysis was required. The Phase 3 SDP analysis, performed with the assistance of the NRC probabilistic risk analysis staff, determined that the resultant Core Damage Frequency and Large Early Release Frequency associated with this finding were less than 1E-6 per year and 1E-7 per year, respectively. Based on these results, this issue was determined to be of very low safety significance. (Section 1R12)
C
: '''Green.'''
The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to take corrective action to preclude the repetition of reactor control instrumentation 24-volt direct current power supply failures. Specifically, the licensee failed to perform weekly verification of control group power supplies to ensure that the "power available" status lights were lit. This corrective action was identified by the licensee in response to the Unit 2 reactor trip on May 12, 2002, which was caused by the failure of redundant power supplies in reactor control instrumentation cabinet 2-PS-CGC-16. The licensee subsequently performed this check on November 22, 2002, and discovered a failed 24-volt direct current power supply in Unit 1 cabinet 1-PS-CGC-16.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this issue could be reasonably viewed as a precursor to a significant event (i.e., potentially result in a reactor trip similar to the Unit 2 trip on May 12, 2002), and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective.
 
Specifically, the failure of redundant power supplies in reactor control instrumentation cabinets would upset plant stability (cause a reactor trip) and challenge the function of critical safety equipment. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding contributes to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available, the inspectors determined that this finding required a Phase 2 SDP analysis. After a review of additional information, the inspectors determined that a Phase 3 analysis was required.
 
The Phase 3 SDP analysis, performed with the assistance of the NRC probabilistic risk analysis staff, determined that the resultant Core Damage Frequency and Large Early Release Frequency associated with this finding were less than 1E-6 per year and 1E-7 per year, respectively. Based on these results, this issue was determined to be of very low safety significance. (Section 1R12)
C
: '''Green.'''
A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V,
  "Instructions, Procedures, and Drawings," was self-revealed. The licensee failed to provide an appropriate procedure for testing the Unit 1 pressurizer power operated relief valves (PORVs), causing an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank. This issue was self-revealed on June 5, 2002, when pressurizer PORV 1-NRV-153 inadvertently opened while testing actuation logic circuitry for pressurizer PORV 1-NRV-151.
 
The surveillance test procedure failed to provide adequate control of 1-NRV-151 and 1-NRV-153, which have a common automatic opening signal. The release rate exceeded the 25 gallons-per-minute limit established for declaring an Unusual Event in accordance with the licensees Emergency Plan.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this issue could be reasonably viewed as a precursor to a significant event and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective. Specifically, the uncontrolled release of reactor coolant system inventory upset plant stability and challenged the inventory control safety function. Because Unit 1 was in a shutdown mode during this period, the inspectors performed a Phase 1 SDP review of this issue using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." Based on the plant conditions at the time, the inspectors concluded that the most appropriate Appendix G checklist to use for this issue was the checklist for "Pressurized Water Reactor Hot Shutdown Operation - Time to core boiling less than 2 hours." Because, operator intervention was required to manually close the affected PORV block valve, the inspectors concluded that the unit was in a configuration where a single active failure or personnel error could have resulted in a rapid loss of reactor coolant system inventory as described in Section II.B.(2) of the checklist.
 
Consequently, the inspectors concluded that this issue increased the likelihood of a loss of reactor coolant system inventory and therefore required a Phase 2 SDP analysis. The inspectors discussed the safety significance of this issue with the Regional Senior Reactor Analyst (SRA). The SRA reviewed the finding and determined that the drain path could be easily isolated, accurate reactor coolant system level indication was available, all steam generators were available for cooling, and all trains of standby injection were available and not impacted by the finding. Based on these factors the finding was determined to be of very low safety significance. (Section 4OA3.1)
C
: '''Green.'''
The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to provide appropriate instructions for conducting a planned shutdown of Unit 2 on January 19, 2002, which resulted in unnecessarily challenging the automatic start function of Unit 2 turbine driven auxiliary feedwater pump (TDAFWP). This issue was self-revealed when the TDAFWP unexpectedly started due to low steam generator levels following the manual reactor trip.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective and was therefore more than a minor concern. Specifically, the function of critical safety equipment was challenged and plant stability was upset during the performance of a normal plant shutdown by the automatic start of Unit 2 TDAFWP. The inspectors performed a Phase 1 SDP review of this issue using the guidance provided in NRC Inspection Manual Chapter 0609,
Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding did not cause or contribute to the likelihood of an initiating event, the inspectors concluded that this issue was of very low safety significance. (Section 4OA3.3)
 
===Cornerstone: Mitigating Systems===
 
C
: '''Green.'''
The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to assure that corrective actions were taken to preclude repetition of emergency diesel generator (EDG) starting air system relay failures, a significant condition adverse to quality. This issue was self-revealed when the failure of a starting air system relay for the Unit 2 AB EDG occurred on October 16, 2002, causing the engine to roll without a valid start signal. The inspectors subsequently identified that appropriate corrective actions to prevent repetition had not been taken following two previous age-related EDG air start relay failures in January 1999 and September 2000.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this issue, if left uncorrected, would become a more significant safety concern and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the mitigating systems cornerstone and adversely affected the cornerstone objective. Specifically, the repetitive EDG air start relay failures affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined that this finding was a licensee performance deficiency of very low safety significance because the finding: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time; (4) did not represent an actual loss of safety function of one or more Non-Technical Specification trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. (Section 4OA2.1)
 
===Cornerstone: Barrier Integrity===
 
C
: '''Green.'''
The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to identify and take appropriate corrective actions to preclude the failure of four Unit 1 reactor coolant system pressure boundary charging line check valves (Velan Model B10-3114B-13M), which were at risk of common cause failure due to industry identified design and manufacturing defects, a significant condition adverse to quality. This issue was self-revealed when the check valves were all found to be stuck in either the full or partially open position during radiographic nonintrusive testing in May 2002.
 
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this finding was associated with the barrier integrity cornerstone and adversely affected the cornerstone objective, and as such it was more than a minor concern. Specifically, the charging line check valves perform a safety-related function of limiting the release of reactor coolant inventory should a charging line failure occur. The failure of the valves in the open position would prohibit the performance of this function and therefore affects the objective of the barrier integrity cornerstone.
 
The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A,
        "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding involved the integrity of the reactor coolant system barrier, the inspectors determined that this finding required a Phase 2 SDP analysis. After consulting with the Regional Senior Reactor Analyst, the inspectors determined that this issue was of very low safety significance because no actual loss of safety function occurred. The inspectors concluded that no actual loss of safety function occurred based on the reported minimal force required to shut the valves (indicating they would have shut given the differential pressure applied during accident conditions) and the redundancy provided by a third check valve (1-CS-321) in the charging line. In accordance with IMC 0609,
Appendix A, Attachment 1, Step 2.6, the SDP results were not evaluated for potential risk contribution due to Large Early Release Frequency because the accident sequence result was less than 1E-7 per year. (Section 4OA2.2)
 
===Licensee Identified Violations===
 
None


=REPORT DETAILS=
=REPORT DETAILS=
===Summary of Plant Status===
Unit 1 operated at or near full power during this inspection period with the following exceptions:
C      On October 5, 2002, the licensee reduced power to approximately 90 percent of rated thermal power to repair a steam leak on a feedwater heater. Following the repair, the licensee returned the unit to full power on October 7, 2002.
C      On November 10, 2002, the licensee initiated a power reduction to approximately 30 percent of rated thermal power to enter the Containment Building and add oil to a reactor coolant pump motor. Following the maintenance activity, the licensee returned the unit to full power on November 12, 2002.
C      On December 21, 2002, the licensee initiated a power reduction to approximately 53 percent of rated thermal power to enter the Containment Building and add oil to a reactor coolant pump motor. Following the maintenance activity, the licensee returned the unit to full power on December 22, 2002.
C      On December 24, 2002, the licensee reduced power to approximately 55 percent of rated thermal power to remove a main feedwater pump from service and repair a failed weld on a small diameter instrument line at the discharge of the pump. Following the repair, the licensee returned the unit to full power on December 25, 2002.
Unit 2 operated at or near full power during this inspection period.
C      On November 4, 2002, the licensee received approval of a Notice of Enforcement Discretion to extend the 72-hour allowed action time of Technical Specification 3.8.1.1.b to preclude shutting down the unit until the CD emergency diesel generator could be restored to an operable status.
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency        Preparedness
{{a|1R01}}
==1R01 Adverse Weather Protection==
{{IP sample|IP=IP 71111.01}}
====a. Inspection Scope====
The inspectors reviewed the licensees procedures and preparations for cold weather conditions. The inspectors reviewed winterization procedures, severe weather procedures, emergency plan implementing procedures related to severe weather, and performed general area walkdowns. Specifically:
C      During general pre-winterization walkdowns conducted the week of October 14, 2002, the inspectors toured selected buildings and areas to verify the licensee had identified all discrepant conditions such as damaged doors, windows, or vent louvers. Additionally, the inspectors observed housekeeping conditions and verified that materials capable of becoming airborne missile hazards during high wind conditions, or impacting snow removal, were appropriately located and restrained.
C        During post-winterization walkdowns conducted the week of November 18, 2002, the inspectors verified that all items on the licensees pre-winterization checklist were completed with appropriate corrective actions taken for identified discrepant conditions. Additionally, the inspectors verified that outside water storage tanks (e.g., refueling water storage tanks, primary water storage tanks, and condensate storage tanks) and associated valve houses and piping had no missing or damaged insulation and were serviced by operable heat trace circuits.
C        During post-winterization walkdowns conducted the week of December 9, 2002, the inspectors toured plant areas to monitor the physical condition of cold weather protection features following a period of extended freezing temperatures. The inspectors observed insulation, heat trace circuits, space heater operation, and weatherized enclosures to ensure operability of affected systems.
====b. Findings====
No findings of significance were identified. {{a|1R04}}
==1R04 Equipment Alignment==
{{IP sample|IP=IP 71111.04}}
===.1 Partial System Walkdowns===
====a. Inspection Scope====
The inspectors performed partial system walkdowns of the following risk-significant systems:
Initiating Events Cornerstone C        Unit 2 AB Emergency Diesel Generator (EDG)
Mitigating Systems Cornerstone C        Unit 2 Turbine Driven and West Auxiliary Feedwater (AFW) System Trains C        Unit 1 West Essential Service Water (ESW) System Train The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones. The inspectors reviewed operating procedures, Technical Specification (TS) requirements, Administrative Technical Requirements, system diagrams, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components were aligned correctly.
In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for equipment alignment related issues documented in selected condition reports (CRs).
====b. Findings====
No findings of significance were identified. {{a|1R05}}
==1R05 Fire Protection==
{{IP sample|IP=IP 71111.05}}
===.1 Routine Resident Inspector Tours===
====a. Inspection Scope====
The inspectors performed fire protection walkdowns of the following risk-significant plant areas:
Initiating Events Cornerstone C        Lake Screen House (Zone 142)
Mitigating Systems Cornerstone C        Fire Pump Building C        Auxiliary Building North 609 Foot Elevation (Zone 44N)
C        Auxiliary Building South 609 Foot Elevation (Zone 44S)
C        Unit 1 East ESW Pump Room (Zone 29A)
C        Unit 1 West ESW Pump Room (Zone 29B)
C        Unit 1 Turbine Building Southeast (Zone 91)
C        Unit 2 Turbine Building Northeast (Zone 96)
C        Unit 1 Turbine Building Southwest (Zone 92)
C        Unit 2 Turbine Building Northwest (Zone 99)
C        Unit 1 CD EDG Room (Zone 15)
C        Unit 2 CD EDG Room (Zone 18)
The inspectors verified that fire zone conditions were consistent with assumptions in the licensees Fire Hazard
=====Analysis.=====
The inspectors walked down fire detection and suppression equipment, assessed the material condition of fire control equipment, and evaluated the control of transient combustible materials.
====b. Findings====
No findings of significance were identified. {{a|1R11}}
==1R11 Licensed Operator Requalification==
{{IP sample|IP=IP 71111.11}}
===.1 Resident Inspector Quarterly Review===
====a. Inspection Scope====
The inspectors assessed licensed operator performance and the training evaluators critique during licensed operator annual requalification evaluations in the D. C. Cook Plant operations training simulator on October 29, 2002. The inspectors focused on alarm response, command and control of crew activities, communication practices, procedural adherence, and implementation of emergency plan requirements.
====b. Findings====
No findings of significance were identified. {{a|1R12}}
==1R12 Maintenance Effectiveness==
{{IP sample|IP=IP 71111.12}}
====a. Inspection Scope====
The inspectors evaluated the licensees handling of selected degraded performance issues involving the following risk-significant structures, systems, and components (SSCs):
Initiating Events Cornerstone C        Control Group Power Supply Failures The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the SSCs. Specifically, the inspectors independently verified the licensees management of handling of SSC performance or condition problems in terms of:
C        appropriate work practices, C        identifying and addressing common cause failures, C        scoping of SSCs in accordance with 10 CFR 50.65(b),
C        characterizing SSC reliability issues, C        tracking SSC unavailability, C        trending key parameters (condition monitoring),
C        10 CFR 50.65(a)(1) or (a)(2) classification and reclassification, and C        appropriateness of performance criteria for SSCs/functions classified (a)(2)and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified (a)(1).
====b. Findings====
b.1 Failure to Address Age-related Failures of Reactor Control Instrumentation Power Supplies to Prevent Repetition of Power Supply Failures The inspectors identified a finding of very low safety significance (Green) associated with a self-revealed event. The licensee failed to assure that prompt corrective actions were taken to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures, a significant condition adverse to quality. The inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned this finding as a Non-Cited Violation.
Discussion On May 12, 2002, an automatic reactor trip of Unit 2 occurred due to the failure of redundant 24-volt direct current (DC) power supplies in reactor control instrumentation cabinet 2-PS-CGC-16. The failure of both power supplies caused the number 21 steam generator feedwater regulating valve to close. Unit 2 subsequently tripped on low steam generator water level coincident with low feedwater flow.
Each reactor control instrumentation cabinet contains two separate power supplies (originally Lambda Model LRS-57-24 or LMS-9120). The two power supplies are interconnected through auctioneering diodes, such that the cabinet remains energized in the event of the failure of one of the power supplies. The cabinets provide indication and control functions for various plant systems including: steam generator feedwater control, automatic steam generator power operated relief valve (PORV) control, automatic steam dump control, reactor coolant system volume control tank automatic make-up, automatic switch-over of the charging pump suction to the refueling water storage tank on low-low volume control tank level, automatic pressurizer pressure control using spray valves and heaters, automatic pressurizer level control, and automatic pressurizer PORV controls. Detection of a single power supply failure was inhibited because there was no annunciation on the loss of a single power supply.
There were "power available" status lights connected with each of the power supplies located inside the normally closed cabinet doors; however, the licensee did not routinely check the status lights prior to the Unit 2 reactor trip.
The reactor control instrumentation cabinet power supplies in question were originally installed in both units in 1994 as part of a modification to replace obsolete equipment.
In 1999 and 2000, several of these power supplies failed and were sent to a vendor for repair. Repair reports were generated by the vendor which identified the existence of internal components that were much older than expected. However, these repair reports were apparently not forwarded to the system engineering department when they were received at D. C. Cook. Following the Unit 2 reactor trip, the two failed power supplies from 2-PS-CGC-16 and several other failed 24-volt DC power supplies were sent to the vendor for detailed analysis. All of the failures were determined to be age-related. In all cases, capacitors with date codes as early as 1989 were found.
Hence, these power supplies were already several years old when they were first installed and energized.
The licensee recognized in August 2001 that there had been a significant number of DC power supply failures during the 24-month period prior to August 2001. The licensee collectively documented a total of 20 power supply failures in CR 01236037, including six reactor control instrumentation power supply failures, stating that the failures should be investigated for a common cause. Other power supply failures were in nuclear instrumentation, radiation monitoring instrumentation, reactor protection instrumentation, rod control/rod position indication, steam generator PORV indication, reactor coolant pump vibration monitoring instrumentation, and main generator hydrogen and carbon dioxide (CO2) purity monitoring. The inspectors noted that the licensee did not complete its evaluation of CR 01236037 until after the Unit 2 reactor trip 9 months later.
It is also noteworthy that Unit 2 was started up following the Cycle 13 refueling outage in February 2002, with one of the two power supplies known to be failed in reactor control instrumentation cabinet 2-PS-CGC-19. This failed power supply was discovered by instrument technicians during a routine cleaning and inspection of the cabinet on February 16, 2002, (12 days prior to completion of the outage). According to the licensees root cause evaluation, replacement of the power supply was not performed due to perceived time pressure associated with the refueling outage schedule.
Considering that a second power supply failure in that cabinet would result in a reactor trip and that the licensee should have been aware of the power supply history based on CR 01236037, the inspectors concluded that this decision was not conservative in that it increased the likelihood an initiating event (i.e., a reactor trip).
The inspectors determined that the licensees failure to assure that corrective actions were taken to preclude repetitive age-related failures of reactor control instrumentation power supplies is a licensee performance deficiency warranting a significance evaluation. The inspectors also concluded that this finding affected the cross-cutting issue of problem identification and resolution.
Analysis The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors concluded that this issue, if left uncorrected, would become a more significant safety concern with the likelihood of continued failures of reactor control instrumentation power supplies and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective. Specifically, the failure of redundant power supplies in reactor control instrumentation cabinets would upset plant stability (cause a reactor trip) and challenge the function of critical safety equipment.
The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding contributes to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available, the inspectors determined that this finding required a Phase 2 SDP analysis.
Using the current risk-informed inspection notebook for D. C. Cook (Revision 0) for the Phase 2 SDP analysis, the inspectors determined that this finding was potentially greater than very low safety significance. Specifically, the inspectors determined that this issue caused the likelihood for transients involving a loss of the primary conversion system (PCS) to be increased by an order of magnitude using Usage Rule 1.2 of IMC 0609, Appendix A, Attachment 2. The initiating event likelihood was evaluated for a greater than 30-day period because the condition had existed since the last refueling outage which occurred in February 2002. However, after a review of additional information the inspectors determined that a Phase 3 analysis was required.
Specifically, for transient sequences, including those involving a loss of the PCS, the inspectors noted that additional credit that was not assumed in the risk-informed notebook could be given to the AFW function. This additional credit involves the operators ability to cross-tie the opposite units motor-driven AFW pumps to the affected units AFW system. This cross-tie evolution is probabilistically limited to the operators ability to perform the evolution and is given a failure probability of 0.1. Applying this additional credit (i.e., one point), the inspectors determined that this finding was of very low safety significance from a Core Damage Frequency (CDF) perspective.
The inspectors also evaluated the effect of this finding on the Large Early Release Frequency (LERF) while factoring in the additional AFW credit discussed above. Using IMC 0609, Appendix H, "Containment Integrity SDP," the inspectors determined that this finding was also potentially greater than very low safety significance. Specifically, for ice condenser plants involving transient accident sequences, the LERF result is a direct correlation to the CDF result. However, the inspectors determined through discussions with NRC risk analysts that due to refinements of LERF estimations, the impact from transient sequences were no longer being considered as a direct correlation to the LERF result. In fact, this refinement indicates that the LERF contribution from most transient sequences for ice condenser plants is not risk significant. The refinement indicates that only station blackout accident sequences would result in this one-to-one correlation between the CDF result and the LERF estimation. When considering this refinement for LERF estimations, the inspectors determined that this finding was also of very low safety significance from a LERF perspective.
Enforcement 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that 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. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the above, the licensee failed to promptly take corrective action to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures, a significant condition adverse to quality. Specifically, for a 24-month period prior to August 2001, the licensee documented six reactor control instrumentation power supply failures. All of these failures were subsequently determined to be age-related.
Consequently, four additional reactor control instrumentation power supply failures have occurred for the same cause since August 2001:
: (1) two redundant power supplies failed in reactor control instrumentation cabinet 2-PS-CGC-16, which resulted in a reactor trip and challenged the function of critical safety equipment;
: (2) one power supply failed in reactor control instrumentation cabinet 2-PS-CGC-19; and
: (3) one power supply failed in reactor control instrumentation cabinet 1-PS-CGC-16. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-316-02-09-01(DRP)). The licensee entered this violation into its corrective action program as CR 02133001 and CR 02133002.
b.2 Failure to Implement a Corrective Action to Prevent Recurrence Associated with Reactor Control Instrumentation Power Supply Failures The inspectors identified a finding of very low safety significance (Green). The licensee failed to take corrective action in response to a Unit 2 reactor trip on May 12, 2002, to preclude the repetition of reactor control instrumentation 24-volt DC power supply failures, a significant condition adverse to quality. Specifically, the licensee did not perform prescribed weekly verifications of reactor control instrumentation power supplies to identify failed power supplies in lieu of no annunciation on the loss of a single power supply. The inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned this finding as a Non-Cited Violation.
Discussion Following the Unit 2 reactor trip on May 12, 2002, the licensee initiated the following corrective actions to prevent recurrence:
C        The failed 24-volt DC power supplies in reactor control instrumentation cabinet 2-PS-CGC-16 and 2-PS-CGC-19 were replaced.
C        All 24-volt DC control group power supplies in Unit 2 were inspected and components were verified to be no older than 2 years old. One power supply was replaced as a result of the inspection.
C        All 24-volt DC control group power supplies in Unit 1 were replaced prior to the Unit 1 reactor startup on June 8, 2002.
C        The licensee identified the need to establish a recurring task to perform a weekly verification of control group power supplies to ensure that the "power available" status lights were lit. This could afford the licensee an opportunity to take compensatory measures or replace a failed power supply prior to the failure of the redundant power supply.
During the inspectors review of the licensees corrective actions for the Unit 2 reactor trip and in response to the inspectors questions, the licensee discovered that weekly verifications of the control group power supply "power available" status lights were not being performed. Verification of the power supply status lights was stipulated as a restart action by the licensees Plant Operations Review Committee following the reactor trip and was specified as a corrective action in the Licensee Event Report (LER) that reported the event. The licensee subsequently performed this check on November 22, 2002, and discovered a failed 24-volt DC power supply in Unit 1 cabinet 1-PS-CGC-16.
The inspectors concluded that the licensees failure to implement this corrective action to prevent recurrence for a significant condition adverse to quality was a performance deficiency warranting a significance evaluation. The inspectors also concluded that this finding affected the cross-cutting issue of problem identification and resolution.
Analysis The inspectors assessed this finding using the SDP. The inspectors concluded that this issue could be reasonably viewed as a precursor to a significant event (i.e., potentially result in a reactor trip similar to the Unit 2 trip on May 12, 2002), and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective.
Specifically, the failure of redundant power supplies in reactor control instrumentation cabinets could upset plant stability (i.e., cause a reactor trip) and challenge the function of critical safety equipment. Consistent with the SDP evaluation performed for the finding described in Section 1R12.b.1, this finding was determined to be of very low safety significance.
Enforcement 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that 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. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the above, the licensee failed to take corrective action to preclude the repetition of reactor control instrumentation 24-volt DC power supply failures. Specifically, the licensee failed to perform a weekly verification of control group power supplies to ensure that the "power available" status lights were lit. This corrective action was identified by the licensee in response to the Unit 2 reactor trip on May 12, 2002, which was caused by the failure of redundant power supplies in reactor control instrumentation cabinet 2-PS-CGC-16. The licensee subsequently performed this check on November 22, 2002, and discovered a failed 24-volt DC power supply in Unit 1 cabinet 1-PS-CGC-16. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-316-02-09-02(DRP)). The licensee entered this violation into its corrective action program as CR 02325058.
{{a|1R13}}
==1R13 Maintenance Risk Assessments and Emergent Work Evaluation==
{{IP sample|IP=IP 71111.13}}
====a. Inspection Scope====
The inspectors reviewed the licensees evaluation and management of plant risk for maintenance activities affecting the following equipment:
Initiating Events Cornerstone C        Unit 1 AB EDG C        Unit 1 AB EDG ESW Supply Valves C        Unit 2 CD EDG Mitigating Systems Cornerstone C        Unit 1 East ESW Pump C        Unit 1 West ESW Pump C        Unit 2 East ESW Pump C        Unit 2 East AFW System Train These activities were selected based on their potential risk significance relative to the reactor safety cornerstones. The maintenance associated with the Unit 2 CD EDG was emergent work to replace the engines governor that was identified as failed during a scheduled surveillance test. As applicable for each of the above activities, the inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensees probabilistic risk analyst and/or shift technical advisor, and verified that plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify that risk analysis assumptions were valid and applicable requirements were met.
In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for maintenance risk related issues that were documented in selected CRs.
====b. Findings====
No findings of significance were identified. {{a|1R14}}
==1R14 Personnel Performance During Non-routine Plant Evolutions==
{{IP sample|IP=IP 71111.14}}
===.1 Unit 1 Power Reduction to Support Oil Addition to a Reactor Coolant Pump Motor===
====a. Inspection Scope====
On November 10, 2002, the licensee initiated a power reduction on Unit 1 to approximately 30 percent of rated thermal power to enter the Containment Building and add oil to the number 14 reactor coolant pump motor. Following the maintenance activity, the licensee returned the unit to full power on November 12, 2002. The inspectors observed portions of the power reduction and assessed operator performance.
====b. Findings====
No findings of significance were identified.
===.2 Unit 1 Control Group Power Supply Replacement===
====a. Inspection Scope====
On December 3, 2002, the licensee replaced one of two redundant 24-volt DC power supplies in reactor control instrumentation cabinet 1-PS-CGC-16. The licensee identified the failed power supply on November 22, 2002 and installed a temporary back-up power supply until replacement of the failed power supply could be performed.
This was the first time that replacement of a control group power supply was performed with the unit on line. The inspectors reviewed the licensees preparations for this evolution and assessed operator performance when the temporary back-up power supply unexpectedly failed, causing a complete loss of power to the instrumentation cabinet.
====b. Findings====
No findings of significance were identified. {{a|1R15}}
==1R15 Operability Evaluations==
{{IP sample|IP=IP 71111.15}}
====a. Inspection Scope====
The inspectors reviewed the following CRs to ensure that either:
: (1) the condition did not render the involved equipment inoperable or result in an unrecognized increase in plant risk, or
: (2) the licensee appropriately applied TS limitations and appropriately returned the affected equipment to an operable status.
Mitigating Systems Cornerstone C        CR 02136008            Wire Found Disconnected in 1-RPS-A C        CR 02131018            Review Operability and Reportability Issues for Two Items Dealing with Feedwater Pressure Indication and the Plant Process Computer Calorimetric Program C        CR 02290012            Steam Generator PORV Actuator Capability Calculation Revealed Negative Calculated Margin for Full Stroke Capability C        CR 02339016            Ultrasonic Examination on Unit 1 ESW to West Motor Driven AFW Pump Piping Found Some Silt/Sand in the Piping Barrier Integrity Cornerstone C        CR 02135049            1-CCR-462 Leaking Excessively During Local Leak Rate Testing C        CR 02300002              Unit 2 Control Room Access Door 2-DR-AUX411B Latch Has Broken and Door Will Not Shut In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for issues potentially affecting the operability of safety-related SSCs that were documented in selected CRs.
====b. Findings====
No findings of significance were identified. {{a|1R16}}
==1R16 Operator Workarounds==
{{IP sample|IP=IP 71111.16}}
===.1 Review of Selected Operator Workarounds===
====a. Inspection Scope====
The inspectors evaluated the operator work-arounds (OWAs) listed below to identify any potential affect on the functionality of mitigating systems or on the operators response to initiating events:
C        OWA 01-02        Feedwater Preheat Valves Cause Cooldown During a Reactor Trip C        OWA 02-03        Need to Open Condenser Vacuum Breakers in Unit 1 to Control Main Turbine Vibration Post Trip The inspectors selected OWA 01-02 to review the potential affect that leakby past the feedwater preheat control valves has on contributing to excessive plant cooldowns following reactor trips from low power. The inspectors selected OWA 02-03 to review the potential for loss of the secondary heat sink with operation of the main turbine vacuum breakers. The inspectors interviewed operating and engineering department personnel and reviewed selected procedures and documents.
====b. Findings====
No findings of significance were identified.
===.2 Semiannual Review of the Cumulative Effect of Operator Workarounds===
====a. Inspection Scope====
The inspectors reviewed the cumulative effect of OWAs, control room deficiencies, and degraded conditions on equipment availability, initiating event frequency, and the ability of the operators to implement abnormal or emergency operating procedures. During this review the inspectors considered the cumulative effects of OWAs on the following:
C        the reliability, availability and potential for mis-operation of a system; C      the ability of operators to respond to plant transients or accidents in a correct and timely manner; and C      the potential to increase an initiating event frequency or affect multiple mitigating systems.
In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for issues potentially affecting the functionality of mitigating systems or on the operators response to initiating events that were documented in selected CRs.
====b. Findings====
No findings of significance were identified. {{a|1R19}}
==1R19 Post Maintenance Testing==
{{IP sample|IP=IP 71111.19}}
====a. Inspection Scope====
The inspectors reviewed the post maintenance testing associated with the following scheduled maintenance activities:
Barrier Integrity Cornerstone C      Unit 1 Leak Test of Post Accident Containment Hydrogen Monitoring System Mitigating Systems Cornerstone C      Unit 2 CD EDG Governor Replacement C      Unit 2 East AFW Pump Maintenance C      Unit 2 West ESW Pump Maintenance The inspectors selected these post maintenance testing activities because the systems were identified as risk significant in the licensees risk analysis. The inspectors reviewed the scope of the work performed and evaluated the adequacy of the specified post maintenance testing. The inspectors verified that the post maintenance testing was performed in accordance with approved procedures, that the procedures clearly stated acceptance criteria, and that the acceptance criteria were met. During this inspection, the inspectors interviewed operations, maintenance, and engineering department personnel and reviewed the completed post maintenance testing documentation.
====b. Findings====
No findings of significance were identified. {{a|1R22}}
==1R22 Surveillance Testing==
{{IP sample|IP=IP 71111.22}}
====a. Inspection Scope====
For the surveillance test procedures listed below, the inspectors observed selected portions of the surveillance test and/or reviewed the test results to determine whether risk significant systems and equipment were capable of performing their intended safety functions and to verify that testing was conducted in accordance with applicable procedural and TS requirements:
Barrier Integrity Cornerstone C      01-OHP-4030-STP-011, "Containment Isolation and Inservice Inspection Valve Operability Test" C      02-IHP-4030-234-001, "Unit 2 Distributed Ignition System Surveillance and Baseline Testing" C      12-IHP-4030-046-227, "Unit 1 and 2 Personnel Airlock Door Seal Leak Rate Surveillance" Mitigating Systems Cornerstone C      01-EHP-4030-ATR-225-020, "Unit 1 Auxiliary Cable Vault CO2 Fire Suppression Test" C      02-IHP-4030-SMP-219, "Steam Generator 1 & 2 Steam/Feed Flow Mismatch and Steam Pressure Protection Set I Functional Test and Calibration" C      02-IHP-4030-SMP-222, "Steam Generator 2 & 4 Steam/Feed Flow Mismatch and Steam Pressure Protection Set II Functional Test and Calibration" C      02-IHP-4030-SMP-227, "Steam Pressure Protection Set III Functional Test and Calibration" C      02-IHP-4030-SMP-228, "Steam Pressure Protection Set IV Functional Test and Calibration" C      12-EHP-5030-CAR-001, "Characterization Testing Program" The inspectors reviewed the test methodology and test results in order to verify that equipment performance was consistent with safety analysis and design basis assumptions.
In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for surveillance testing related issues documented in selected CRs.
====b. Findings====
No findings of significance were identified. {{a|1R23}}
==1R23 Temporary Plant Modifications==
{{IP sample|IP=IP 71111.23}}
====a. Inspection Scope====
The inspectors reviewed the temporary modifications listed below to verify that the installations were consistent with design modification documents and that the modifications did not adversely impact system operability or availability:
Barrier Integrity Cornerstone C        12-EHP-5040-EMP-006, "Disable Bridge East Travel Limit Switch on East Auxiliary Building Crane 12-QM-3E" Mitigating Systems Cornerstone C        1-TM-02-85-R0, "Install Backup Power Supply for Control Group 1" C        12-TM-00-61-R2, "Winterization/De-Winterization Temporary Modification to Support 12-IHP-5040-EMP-004" The first temporary modification disabled the end of travel limit switch to allow the Auxiliary Building crane to move far enough to allow resin shipping casks to be lowered into the drumming room. The second temporary modification installed a backup power supply in reactor control instrumentation cabinet 1-PS-CGC-16 until a permanent power supply replacement could be performed for one of the two redundant 24-volt DC power supplies. The third temporary modification installed ventilation system covers and other cold weather system protection measures for the upcoming winter season.
The inspectors verified that configuration control of the modifications were correct by reviewing design modification documents and confirmed that appropriate post-installation testing was accomplished. The inspectors interviewed engineering and operations department personnel and reviewed the design modification documents against the applicable portions of the Updated Final Safety Analysis Report (UFSAR).
====b. Findings====
No findings of significance were identified. {{a|1EP2}}
==1EP2 Alert and Notification System (ANS) Testing==
{{IP sample|IP=IP 71114.02}}
====a. Inspection Scope====
The inspectors discussed with Emergency Preparedness (EP) staff the design, equipment, and periodic testing of the public ANS for the D. C. Cook reactor facility emergency planning zone to verify that the system was properly tested and maintained.
The inspectors also reviewed procedures and records for a 24-month period ending September 2002 related to ANS testing, annual preventive maintenance, and non-scheduled maintenance. The inspectors reviewed the licensees documentation for determining whether each model of siren installed in the emergency planning zone would perform as expected if fully activated. Records used to document and trend component failures for each model of installed siren were also reviewed to ensure that corrective actions were taken for test failures or system anomalies.
====b. Findings====
No findings of significance were identified. {{a|1EP3}}
==1EP3 Emergency Response Organization (ERO) Augmentation Testing==
{{IP sample|IP=IP 71114.03}}
====a. Inspection Scope====
The inspectors reviewed the licensees ERO augmentation testing to verify that the licensee maintained and tested its ability to staff the ERO during an emergency in a timely manner. Specifically, the inspectors reviewed quarterly, off-hours staff augmentation test procedures, dated December 14, 2001, August 23, 2001, March 14, 2002, April 16, 2002, and July 17, 2002 drill records, primary and backup provisions for off-hours notification of the D. C. Cook reactor facility emergency responders, and the current ERO rosters for D. C. Cook. The inspectors reviewed and discussed the facility EP staff's provisions for maintaining ERO call out lists.
====b. Findings====
No findings of significance were identified. {{a|1EP5}}
==1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies==
{{IP sample|IP=IP 71114.05}}
====a. Inspection Scope====
The inspectors reviewed the Performance Assurance staff's 2001 - 2002 audits to ensure that these audits complied with the requirements of 10 CFR 50.54(t) and that the licensee adequately identified and corrected deficiencies. The inspectors also reviewed the EP staff's 2001 and 2002 self-assessments, and critiques to evaluate the EP staff's efforts to identify and correct weaknesses and deficiencies. Additionally, the inspectors reviewed a sample of EP items, CRs, and action requests related to the facility's EP program to determine whether corrective actions were acceptably completed.
====b. Findings====
No findings of significance were identified. {{a|1EP6}}
==1EP6 Drill Evaluation==
{{IP sample|IP=IP 71114.06}}
====a. Inspection Scope====
The inspectors observed the conduct of the licensee's annual announced emergency training exercise that was conducted in the licensee's control room simulator and emergency response facilities on October 23, 2002. The inspection effort was focused on evaluation of the licensee's classifications, notifications, and protective action recommendations for the simulated event. The inspectors also evaluated the licensee's conduct of the training evolution, including the licensees critique of performance to identify weaknesses and deficiencies.
====b. Findings====
No findings of significance were identified.


==RADIATION SAFETY==
==RADIATION SAFETY==
 
===Cornerstone: Occupational Radiation Safety (OS)===
===Cornerstone: Occupational Radiation Safety===
 
2OS1 Access Control to Radiologically Significant Areas (71121.01)
2OS1 Access Control to Radiologically Significant Areas (71121.01)


===.1 Plant Walkdowns, Radiological Boundary Verifications, and Radiation Work Permit===
===.1 Plant Walkdowns, Radiological Boundary Verifications, and Radiation Work Permit===
Reviews
Reviews


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors conducted walkdowns of the radiologically protected area to verify the adequacy of radiological area boundaries and postings. Specifically, the inspectors walked down radiologically significant work area boundaries (radiation, high and locked high radiation areas) in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, as well as the Unit 2 Containment Building. The inspectors performed confirmatory radiation surveys in selected portions of these areas to verify that these areas were properly posted and controlled in accordance with 10 CFR 20, licensee procedures, and TSs. The inspectors also examined the radiological conditions of work areas within those radiation, high and locked high radiation areas to assess the adequacy of licensee implemented contamination controls. Additionally, the inspectors reviewed radiation work permits (RWPs) for general tours, access to locked high radiation areas for work on spent fuel pool demineralizers, drumming room clean-up activities, an at power entry into Unit 1 for work on a reactor coolant pump; and for another at power entry into Unit 2 for work on a safety injection system accumulator. The RWPs were evaluated for protective clothing requirements, respiratory protection concerns, electronic dosimetry alarm set points, use of remote telemetry dosimetry, radiation protection hold points, and As-Low-As-Reasonably-Achievable considerations, to verify that work instructions and controls had been adequately specified and that electronic dosimeter set points were in conformity with survey results.
The inspectors conducted walkdowns of the radiologically protected area to verify the adequacy of radiological area boundaries and postings. Specifically, the inspectors walked down radiologically significant work area boundaries (radiation, high and locked high radiation areas) in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, as well as the Unit 2 Containment. The inspectors performed confirmatory radiation surveys in selected portions of these areas to verify that these areas were properly posted and controlled in accordance with 10 CFR Part 20, licensee procedures, and Technical Specifications. The inspectors also examined the radiological conditions of work areas within those radiation, high and locked high radiation areas to assess the adequacy of licensee implemented contamination controls. Additionally, the inspectors reviewed radiation work permits (RWPs) for general tours, access to locked high radiation areas (LHRAs) for work on spent fuel pool (SFP) demineralizers, drumming room clean-up activities, an at power entry into Unit 1 for work on a reactor coolant pump; and for another at power entry into Unit 2 for work on the no.1 safety injection (SI) accumulator. The RWPs were evaluated for protective clothing requirements, respiratory protection concerns, electronic dosimetry alarm set points, use of remote telemetry dosimetry, radiation protection (RP) hold points, and As-Low-As-Is-Reasonably-Achievable (ALARA) considerations, to verify that work instructions and controls had been adequately specified and that electronic dosimeter set points were in conformity with survey results.


====b. Findings====
====b. Findings====
Line 422: Line 60:


===.2 Job-In-Progress Reviews, Observations of Radiation Worker Performance, and===
===.2 Job-In-Progress Reviews, Observations of Radiation Worker Performance, and===
Radiation Protection Technician Proficiency
Radiation Protection Technician Proficiency


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed selected portions of the following radiologically significant work activities performed during the inspection and evaluated the licensees use of radiological controls:
The inspectors observed selected portions of the following radiologically significant work activities performed during the inspection and evaluated the licensees use of radiological controls:
C        number 21 safety injection system accumulator level indicator repair, and C        preparations for spent fuel pool demineralizer work.
* Unit 2, No.1 SI accumulator level indicator repair; and
 
* Preparations for SFP demineralizer work The inspectors reviewed the pre-job briefing package for the work evolutions, reviewed the radiological requirements for the activities and assessed the licensees performance with respect to those requirements. The inspectors reviewed survey records, including radiation, contamination, and airborne surveys, to verify that appropriate radiological controls were effectively utilized. The inspectors also reviewed in-process surveys and applicable postings and barricades to verify their accuracy. The inspectors observed radiation protection technician (RPT) and worker performance during the work evolution at the job sites to verify that the technicians and workers were aware of the significance of the radiological conditions in their workplace and RWP controls/limits, and that they were performing adequately given the radiological hazards present and the level of their training.
The inspectors reviewed the pre-job briefing package for the work evolutions, reviewed the radiological requirements for the activities and assessed the licensees performance with respect to those requirements. The inspectors reviewed survey records, including radiation, contamination, and airborne surveys, to verify that appropriate radiological controls were effectively utilized. The inspectors also reviewed in-process surveys and applicable postings and barricades to verify their accuracy. The inspectors observed radiation protection technician (RPT) and worker performance during the work evolution at the job sites to verify that the technicians and workers were aware of the significance of the radiological conditions in their workplace and RWP controls/limits, and that they were performing adequately given the radiological hazards present and the level of their training.


====b. Findings====
====b. Findings====
Line 435: Line 71:


===.3 Identification and Resolution of Problems===
===.3 Identification and Resolution of Problems===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed licensee CRs written since the last inspection (July 2002) to the date of the current inspection, which focused on access control to radiologically significant areas (i.e., problems concerning activities in high radiation areas, radiation protection technician performance, and radiation worker practices). The inspectors also reviewed the recently revised "High, Locked High, and Very High Radiation Area Access" procedure, which addressed new requirements for specific locking devices for these areas. The inspectors reviewed these documents to assess the licensee's ability to identify repetitive problems, contributing causes, and the extent of conditions, and then implement corrective actions in order to achieve lasting results.
The inspectors reviewed licensee Condition Reports (CRs) written since the last inspection (July 2002) to the date of the current inspection, which focused on access control to radiologically significant areas (i.e., problems concerning activities in HRAs, radiation protection technicians performance, and radiation worker practices). The inspectors also reviewed the recently revised High, Locked High, and Very High Radiation Area Access procedure, which addressed new requirements for specific locking devices for these areas. The inspectors reviewed these documents to assess the licensees ability to identify repetitive problems, contributing causes, and the extent of conditions, and then implement corrective actions in order to achieve lasting results.


====b. Findings====
====b. Findings====
Line 444: Line 79:


===.1 Walkdowns of Radiation Monitoring Instrumentation===
===.1 Walkdowns of Radiation Monitoring Instrumentation===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the UFSAR and performed walkdowns of continuous air monitors in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, Radioactive Material Building, and one area radiation monitor (ARM) in the Unit 2 Containment Building. Additionally, the inspectors examined a representative number of portable radiation survey instruments staged throughout the licensees facility to verify that those instruments had current calibrations, were operable, and in good physical condition. The inspectors also reviewed the status of repair or troubleshooting activities associated with selected radiation monitoring instruments (i.e., small article monitors and portal monitors that had work request tags) to verify that instrumentation problems were being addressed in an appropriate and timely manner.
The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) and performed walkdowns of continuous air monitors in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, radioactive material building areas, and one area radiation monitor (ARM) in the Unit 2 Containment. Additionally, the inspectors examined a representative number of portable radiation survey instruments staged throughout the licensees facility to verify that those instruments had current calibrations, were operable, and in good physical condition. The inspectors also reviewed the status of repair or troubleshooting activities associated with selected radiation monitoring instruments (i.e. small article monitors and portal monitors that had work request tags)to verify that instrumentation problems were being addressed in an appropriate and timely manner.


====b. Findings====
====b. Findings====
Line 452: Line 86:


===.2 Calibration, Operability, and Alarm Set Points of Radiation Monitoring Instrumentation===
===.2 Calibration, Operability, and Alarm Set Points of Radiation Monitoring Instrumentation===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors examined radiological instrumentation associated with monitoring transient high and/or very high radiation areas to verify that the instrumentation was operating consistent with industry standards and in accordance with station procedures.
The inspectors examined radiological instrumentation associated with monitoring transient high and/or very high radiation areas to verify that the instrumentation was operating consistent with industry standards and in accordance with station procedures.


Specifically, the inspectors assessed the operability of the following instrumentation:
Specifically, the inspectors assessed the operability of the following instrumentation:
C      Unit 2 In-Core Instrumentation Room ARM.
* Unit 2, In-Core Instrumentation Room ARM The inspectors reviewed the licensees alarm set point for this specific ARM to verify that the set point was established consistent with the UFSAR, Technical Specifications, and the stations Emergency Plan.


The inspectors reviewed the licensees alarm set point for this specific ARM to verify that the set point was established consistent with the UFSAR, TSs, and the licensees Emergency Plan.
The inspectors discussed surveillance practices with licensee personnel and reviewed calender year (CY) 2001 - 2002 calibration records and procedures for selected radiation monitors used for assessment of internal exposure. The inspectors also reviewed calibration records and procedures for those instruments utilized for surveys of personnel and equipment prior to egress from the radiologically controlled area (RCA).


The inspectors discussed surveillance practices with licensee personnel and reviewed calender year 2001 - 2002 calibration records and procedures for selected radiation monitors used for assessment of internal exposure. The inspectors also reviewed calibration records and procedures for those instruments utilized for surveys of personnel and equipment prior to egress from the radiologically controlled area. These instruments included:
These instruments included:
C      AMS-4 Air Monitoring System, C      APTEC PMW-3 Personnel Monitor, and C      Gamma 40/60 Portal Monitor.
* AMS-4 Air Monitoring System;
* APTEC PMW-3 Personnel Monitor; and
* Gamma 40/60 Portal Monitor.


Additionally, the alarm set points for these instruments were reviewed to verify that they were established at levels consistent with industry standards and regulatory guidance provided in Health Physics Positions 72 and 250 of NUREG/CR-5569.
Additionally, the alarm set-points for these instruments were reviewed to verify that they were established at levels consistent with industry standards and regulatory guidance provided in Health Physics Positions no. 72 and no. 250 of NUREG/CR-5569.


The inspectors evaluated the calibration procedures and calibration records for selected portable radiation survey instruments to verify that they had been properly calibrated consistent with the licensees procedures. Specifically, the inspectors reviewed the calibrations of the following instruments:
The inspectors evaluated the calibration procedures and CY 2001 - 2002 calibration records for selected portable radiation survey instruments to verify that the had been properly calibrated consistent with the licensees procedures. Specifically, the inspectors reviewed the calibrations of the following instruments:
C      Emergency Plan designated RO-7 ion chamber, and C      Smart Radiation Monitor general area dose rate meter.
* Emergency Plan designated RO-7 ion chamber; and
* Smart Radiation Monitor general area dose rate meter.


The inspectors also assessed periodic performance tests completed for selected portable radiation survey instruments to verify that they had been tested consistent with the licensees procedures. Specifically, the inspectors observed the performance testing of the following instruments:
The inspectors also assessed periodic performance tests (PPTs) completed for selected portable radiation survey instruments to verify that the had been tested consistent with the licensees procedures. Specifically, the inspectors observed the PPTs of the following instruments:
C      extender instruments, and C      bicron RSO survey instruments.
* Extender instruments; and
* RSO instruments.


====b. Findings====
====b. Findings====
Line 476: Line 113:


===.3 Radiation Protection Technician Instrument Use===
===.3 Radiation Protection Technician Instrument Use===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed RPTs performing in-field source checks of portable radiation survey instruments to verify that those source checks were adequately completed using appropriate radiation sources and station procedures. The inspectors assessed the RPTs use of radiation/contamination detection instruments as they provided radiological job coverage for risk significant work (e.g., the safety injection system accumulator repair work in the Unit 2 Containment Building), as well as routine work, to ensure that the RPTs were utilizing the appropriate instruments. The inspectors monitored RPTs performing functional tests of selected contamination monitors, portal monitors, and small article monitors (i.e., for surveys of personnel and equipment prior to unconditional release from the radiologically controlled area) to verify that they were source tested and calibrated as required by station procedures and industry standards.
The inspectors observed RPTs performing in-field source checks of portable radiation survey instruments to verify that those source checks were adequately completed using appropriate radiation sources and station procedures. The inspectors assessed the RPTs use of radiation/contamination detection instruments as they provided radiological job coverage for risk significant work (e.g. the SI accumulator repair work in the Unit 2 Containment), as well as routine work, to ensure that the RPTs were utilizing the appropriate instruments. The inspectors monitored RPTs performing functional tests of selected contamination monitors, portal monitors, and small article monitors (i.e., for surveys of personnel and equipment prior to unconditional release from the RCA) to verify that they were source tested and calibrated as required by station procedures and industry standards.


====b. Findings====
====b. Findings====
Line 484: Line 120:


===.4 Problem Identification and Resolution===
===.4 Problem Identification and Resolution===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed calendar year 2001-2002 CRs that addressed radiation monitoring instrument deficiencies to determine if any significant radiological incidents involving instrument deficiencies had occurred. The inspectors examined the results of a self-assessment (i.e., the Summary Report for performance Assurance Audit PA-0206, "Radiation Protection") that focused on the licensees CR database and several individual CRs related to radiation monitoring instrumentation generated during the current assessment period. The inspectors also interviewed plant staff and examined closed CRs to verify that radiological instrumentation related issues were adequately addressed by the licensee. The inspectors evaluated these documents to verify the licensees ability to identify repetitive problems, contributing causes, extent of conditions, and the implementation of corrective actions to achieve lasting results.
The inspectors reviewed CY 2001-2002 CRs that addressed radiation monitoring instrument deficiencies to determine if any significant radiological incidents involving instrument deficiencies had occurred. The inspectors examined the results of a self-assessment (i.e., the Summary Report for performance Assurance Audit PA-0206, Radiation Protection) that focused on the licensees CR database and several individual CRs related to radiation monitoring instrumentation generated during the current assessment period. The inspectors also interviewed plant staff and examined closed CRs to verify that radiological instrumentation related issues were adequately addressed by the licensee. The inspectors evaluated these documents to verify the licensees ability to identify repetitive problems, contributing causes, extent of conditions, and the implementation of corrective actions to achieve lasting results.
 
====b. Findings====
No findings of significance were identified.
 
==SAFEGUARDS==
 
===Cornerstone: Physical Protection===
 
3PP3 Response to Contingency Events (71130.03)
 
====a. Inspection Scope====
The inspectors reviewed the status of security operations and assessed licensee implementation of the protective measures in place as a result of the current, elevated threat environment.


====b. Findings====
====b. Findings====
Line 505: Line 128:
==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
{{a|4OA1}}
{{a|4OA1}}
==4OA1 Performance Indicator Verification==
==4OA1 Performance Indicator Verification==
{{IP sample|IP=IP 71151}}
{{IP sample|IP=IP 71151}}
===.1 Safety System Functional Failures===
====a. Inspection Scope====
Mitigating Systems Cornerstone The inspectors verified the Safety System Functional Failures performance indicator for both units. The inspectors reviewed each LER from October 2001 to September 2002, determined the number of safety system functional failures that occurred, evaluated each LER against the performance indicator definitions, and verified the number of safety system functional failures reported.
====b. Findings====
No findings of significance were identified.
===.2 Reactor Coolant System Leakage===


====a. Inspection Scope====
====a. Inspection Scope====
Barrier Integrity Cornerstone The inspectors verified the Reactor Coolant System Leakage performance indicator for both units. The inspectors reviewed operating logs and the results of reactor coolant system water inventory balance calculations performed from October 2001 through September 2002 and verified the licensees calculation of reactor coolant system leakage for both units.
The inspectors reviewed the licensees assessment of its performance indicator (PI) for Barrier Integrity, Reactor Coolant System (RCS) Specific Activity. No reportable elements were identified by the licensee for 4th quarter of 2001,and the 1st, 2nd, 3rd quarters of 2002. The inspectors compared the licensees data with CY 2001 and 2002 CRs to verify that there were no occurrences concerning the Barrier Integrity, RCS Specific Activity cornerstone. Additionally, the inspectors also observed staff chemistry technicians collecting RCS samples to verify that the technicians had complied with the applicable procedures during the collection and processing of the samples.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified.
{{a|4OA6}}


===.3 Reactor Coolant System Specific Activity===
====a. Inspection Scope====
Barrier Integrity Cornerstone The inspectors verified the Reactor Coolant System Specific Activity performance indicator for both units. The inspectors reviewed specific activity results reported from October 2001 through September 2002 and verified the licensees calculation of reactor coolant system activity for both units. In addition, the inspectors observed staff chemistry technicians collecting reactor coolant system samples to verify that the technicians had complied with applicable procedures during the collection and processing of the samples.
====b. Findings====
No findings of significance were identified.
===.4 ANS, Drill and Exercise Performance (DEP), and ERO Drill Participation===
====a. Inspection Scope====
Emergency Preparedness Cornerstone The inspectors verified that the licensee had accurately reported the ANS, DEP, and ERO Drill Participation performance indicators for both units. Specifically, the inspectors reviewed the licensees performance indicator records, data reported to the NRC, and CRs for the period July 2001 through September 2002 to identify any occurrences that were not identified by the licensee. Records of relevant control room simulator training sessions, periodic ANS tests, and excerpts of drill and exercise scenarios and evaluations were also reviewed.
====b. Findings====
No findings of significance were identified. {{a|4OA2}}
==4OA2 Identification and Resolution of Problems==
{{IP sample|IP=IP 71152}}
===.1 EDG Starting Air Relay Failures===
====a. Inspection Scope====
A failure of a starting air system relay for the Unit 2 AB EDG occurred on October 16, 2002, causing the engine to roll without a valid start signal. The inspectors reviewed previous failures associated with starting air system relays for the EDGs. In addition, the inspectors reviewed the root cause evaluation for the following CR:
C        CR P-99-01279, "Unit 2 AB EDG Rolled with Air by Itself. An Auxiliary Equipment Operator Was Dispatched Who Reported the Engine Rolling with Air.
No Indication of a Start Signal Was Detected Locally or in the Control Room.
Starting Air Continued to Blow Down Engine Until Air Depleted."
The inspectors verified the following attributes during their review of the licensees corrective actions for the above CR and several other related CRs:
C        consideration of the extent of condition, generic implications, common cause and previous occurrences; C        classification and prioritization of the resolution of the problem, commensurate with safety significance; C        identification of the root and contributing causes of the problem; and C        identification of corrective actions which were appropriately focused to correct the problem.
The inspectors discussed the corrective actions and associated CR evaluations with site personnel including the CR evaluators and system engineers.
====b. Findings====
The inspectors identified a finding of very low safety significance (Green) associated with this self-revealed event. The licensee failed to assure that corrective actions were taken to preclude repetition of EDG starting air system relay failures, a significant condition adverse to quality. The inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned this finding as a Non-Cited Violation.
Description On October 16, 2002, one of the two Unit 2 AB EDG air receivers (AB2) depressurized when starting air valve 2-XRV-222 inadvertently opened and consequently air rolled the engine. The second safety-related air receiver for the EDG remained pressurized and auxiliary equipment operators manually isolated 2-XRV-222, which re-pressurized the AB2 air receiver. A subsequent licensee investigation determined that starting air system relay 2-19-DGAB had failed causing the starting air valve to open.
The starting air system provides the motive force for starting the EDGs. The inspectors noted that each of the four plant EDGs have two of the same model relays installed in the starting air system, which are designated as relays 19 and 19-1. The inspectors determined that while the failure of relay 19 resulted in depressurization of only one of the two air start receivers for an EDG, the failure of relay 19-1 resulted in depressurization of both air start receivers for an EDG.
The inspectors reviewed work order history and the corrective action program database to determine whether or not previous failures had occurred on starting air system relays.
The inspectors found two recent occurrences:
C      in January 1999, relay 19-1 failed on the Unit 2 AB EDG starting air system, which resulted in depressurization of the starting air system; and C      in September 2000, relay 19 failed on the Unit 1 CD EDG starting air system, which resulted in depressurization of one air start receiver.
The inspectors reviewed the licensees assessment of CR P-99-01279 associated with the January 1999 failure and noted the following:
C      the root cause evaluation concluded that the most likely failure scenario of the relay was long term overheating of the continuously energized coil in relay 19-1; C      the root cause evaluation concluded that while the relay was designed and rated for 250-volt DC, equalization of the station batteries at the plant had raised DC bus voltages to 280-volt DC. The evaluator determined that from a thermal deterioration perspective, this would increase the heat generated in the relay by approximately 25 percent; C      the root cause evaluation concluded that given the root cause of the failure, the population of similarly aged relays that were continuously energized should be considered suspect and candidates for failure; C      the root cause evaluation concluded that similar relays should be expected to function for an equivalent duration given similar operating conditions; and C      the root cause evaluation recognized that the recommendations in the evaluation were based on only one data point and that prior to a wholesale changeout of relays, the analysis of additional relays of the same type in the same configuration should be considered.
The inspectors determined that the licensees root cause evaluation addressed the potential extent of condition and the generic implications of the relay failure. However, the only corrective action taken for this significant condition adverse to quality was the replacement of the failed relay 19-1 for the Unit 2 AB EDG. The licensee did not evaluate the condition of other relays of the same type in the same configuration, nor investigate the need for the implementation of a preventive maintenance program or replacement program for this type of relay used in the EDG starting air system.
The inspectors also reviewed CR 00266004, which documented the failure of relay 19 for the Unit 1 CD EDG in September 2000. The inspectors noted that neither an apparent cause evaluation nor a root cause evaluation was performed and that the licensee did not evaluate the potential extent of condition considering the previous failure in January 1999. The licensee concluded that the failure was age related and the only corrective action completed was replacement of the failed relay.
The inspectors determined that the licensees failure to assure that corrective actions were taken to preclude repetition of starting air system relay failures for the EDGs is a licensee performance deficiency warranting a significance evaluation. The inspectors also concluded that this finding affected the cross-cutting issue of problem identification and resolution.
Analysis The inspectors assessed this finding using the SDP. The inspectors concluded that this issue, if left uncorrected, would become a more significant safety concern with the continued failures of the starting air system relays and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the mitigating systems cornerstone and adversely affected the cornerstone objective.
Specifically, the repetitive EDG air start relay failures affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"
and determined that this finding was a licensee performance deficiency of very low safety significance because the finding:
: (1) was not a design or qualification deficiency;
: (2) did not represent an actual loss of safety function of a system;
: (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time;
: (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and
: (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Enforcement 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that 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. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the above, the licensee failed to take corrective action to prevent repetitive failures of EDG starting air system relays 19 and 19-1, a significant condition adverse to quality. Specifically, in January 1999, relay 19-1 failed on the Unit 2 AB EDG starting air system, which resulted in depressurization of the starting air system. The licensee failed to promptly perform corrective actions to preclude a repetition of starting air system relay failures for the EDGs. Consequently, two additional starting air system relay failures have occurred:
: (1) in September 2000, relay 19 failed on the Unit 1 CD EDG starting air system, which resulted in depressurization of one air start receiver; and
: (2) in October 2002, relay 19 failed on the Unit 2 AB EDG starting air system, which resulted in depressurization of one air start receiver. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-315/316-02-09-03(DRP)). The licensee entered this violation into its corrective action program as CR 02289033.
===.2 Common Cause Failure of Four Unit 1 Charging System Check Valves===
====a. Inspection Scope====
During the Unit 1 Cycle 18 refueling outage in May 2002, the licensee discovered the failure of four reactor coolant system pressure boundary charging line check valves.
The inspectors reviewed the circumstances relating to the common cause failure of these valves documented in the root cause evaluation for the following CR:
C      CR 02134021, "Check Valves 1-CS-328-L1, 1-CS-328-L4, 1-CS-329-L1, and 1-CS-329-L4 Were Found Open During Radiographic Nonintrusive Testing."
The inspectors verified the following attributes during their review of the licensees corrective actions for the above CR and several other related CRs:
C      consideration of the extent of condition, generic implications, common cause, and previous opportunities to identify and correct the condition; C      classification and prioritization of the resolution of the problem, commensurate with safety significance; C      identification of the root and contributing causes of the problem; and C      identification of corrective actions which were appropriately focused to correct the problem.
The inspectors also reviewed the corrective actions and associated CR evaluations with applicable site personnel including the CR evaluators and system engineers.
====b. Findings====
The inspectors identified a finding of very low safety significance (Green) associated with this self-revealed event. The licensee failed to identify and take appropriate corrective actions to preclude the failure of reactor coolant system pressure boundary charging line check valves (Velan model B10-3114B-13M), which were at risk of common cause failure due to industry identified design and manufacturing defects, a significant condition adverse to quality. The inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned this finding as a Non-Cited Violation.
Description On May 12, 2002, the licensee conducted radiographic nonintrusive testing on check valve 1-CS-329-L1 in accordance with procedure 12-EP-4030-001-001, "Check Valve Examination Surveillance." Results of the examination indicated that the valves disk was stuck in the open position. Subsequent radiography of the remaining three identical charging line check valves, 1-CS-329-L4, 1-CS-328-L1, and 1-CS-328-L4, identified that all three were stuck in either the full or partially open position. The licensees examination of the valves identified several as-found condition discrepancies including:
C        discs binding against the valve body due to oversize discs; C        improper bushing positioning; and C        problems with disc to arm clearances.
The licensee documented the condition and entered it into its corrective action program as CR 02134021. In response to the CR, the licensee conducted a thorough root cause analysis, which the inspectors concluded properly identified substantial industry operating experience documenting similar failures. Further, the licensees evaluation identified instances of D. C. Cook operating experience involving similar failures which were not considered when evaluating operating experience. Specifically:
C        The Velan Valve Corporation published Service Bulletin 104 on October 10, 1990. The licensee received the bulletin on October 12, 1990 and entered it into the its corrective action program as problem report 90-1503. The licensee took corrective action including one-time vendor training on the identified valve deficiencies and the return of on-site spare parts to the vendor for examination.
No positive actions, however, such as specific dimensional checks, were taken to determine if the condition existed in the in-service components.
C        On February 10, 1992, the licensee received correspondence from Westinghouse Electric Corporation informing them of a 10 CFR Part 21 Report filed by Velan. The correspondence was entered into the licensees corrective action program as problem report 92-157. The issue was closed with no action taken based on the previous evaluation of problem report 90-1503.
C        On January 24, 1996, Operating Experience 7640 was received by the licensee which documented Sequoias discovery of four Velan Model B10-3114B-13MS 3-inch charging injection check valves in the stuck open position. The OE was entered into the licensees corrective action program as CR 96-0094. The licensees evaluation concluded that although similar valves were installed at D. C. Cook, similar problems had not been observed and therefore no corrective action was required. Again, no specific dimensional checks of the installed valves were performed.
The inspectors determined that the licensees failure to assure that corrective actions were taken to preclude the failure of reactor coolant system pressure boundary charging line check valves due to industry identified design and manufacturing defects was a licensee performance deficiency warranting a significance evaluation. The inspectors also concluded that this finding affected the cross-cutting issue of problem identification and resolution. The inspectors noted that the majority of the instances of missed opportunities to identify and correct potential valve deficiencies occurred in the early to mid 1990s time frame.
Analysis The inspectors assessed this finding using the SDP. The inspectors concluded that this finding was associated with the barrier integrity cornerstone and adversely affected the cornerstone objective, and as such it was more than a minor concern. Specifically, the charging line check valves perform a safety-related function of limiting the release of reactor coolant inventory should a charging line failure occur. The failure of the valves in the open position would prohibit the performance of this function and therefore affects the objective of the barrier integrity cornerstone. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding involved the integrity of the reactor coolant system pressure boundary, the inspectors determined that this finding required a Phase 2 SDP analysis. After consulting with the Regional SRA, the inspectors determined that this issue was of very low safety significance because no actual loss of safety function occurred. The inspectors concluded that no actual loss of safety function occurred based on the reported minimal force required to shut the valves (indicating that they would have shut given the differential pressure applied during accident conditions)and the redundancy provided by a third check valve (1-CS-321) in the charging line. In accordance with IMC 0609, Appendix A, Attachment 1, Step 2.6, the SDP results were not evaluated for potential risk contribution due to LERF because the accident sequence result was less than 1E-7 per year.
Enforcement 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that 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. Contrary to the above, the licensee failed to take corrective action to preclude the failure of reactor coolant system pressure boundary charging line check valves (Velan model B10-3114B-13M), which were at risk of common cause failure due to industry identified design and manufacturing defects, a condition adverse to quality. Specifically, industry operating experience was published in October 1990, February 1992, and January 1996 and subsequently entered into the licensees corrective action program. However, the licensee took no positive actions, such as specific dimensional checks, to determine if the condition existed in the in-service components. Consequently, during the Unit 1 Cycle 18 refueling outage in May 2002, the licensee discovered the failure of four reactor coolant system pressure boundary charging line check valves (1-CS-329-L1, 1-CS-329-L4, 1-CS-328-L1, and 1-CS-328-L4), which were the result of conditions identified in the industry operating experience. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-315-02-09-04(DRP)). The licensee entered this violation into its corrective action program as CR 02134021.
{{a|4OA3}}
==4OA3 Event Follow-up==
{{IP sample|IP=IP 71153}}
===.1 (Closed) Unresolved Item (URI) 50-315-02-06-01(DRP): "Pressurizer Power Operated===
Relief Valve (PORV) Inadvertently Opened During Testing Resulting in a Loss of      Reactor Coolant System Inventory and an Unusual Event."
====a. Inspection Scope====
On June 5, 2002, with Unit 1 in Mode 4 (Hot Shutdown), pressurizer PORV 1-NRV-153 inadvertently opened while testing actuation logic circuitry for pressurizer PORV 1-NRV-151. Approximately 100 gallons of reactor coolant was released to the pressurizer relief tank. The release rate exceeded the 25 gallons-per-minute (gpm) limit established for declaring an Unusual Event in accordance with the licensees Emergency Plan. The inspectors reviewed the circumstances associated with this event, including the root cause determination, operator response during the event, and corrective actions.
====b. Findings====
A finding of very low safety significance (Green) was self-revealed. The licensee failed to provide an appropriate procedure for testing the Unit 1 pressurizer PORVs, causing an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank.
This finding was dispositioned as a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Discussion As discussed in NRC Inspection Report 50-315/316-02-06(DRP), the inspectors originally documented this finding as an Unresolved Item pending a final safety significance determination. The inspectors referred this finding to the Regional SRA to perform the additional analysis.
Analysis The inspectors assessed this finding using the SDP. The inspectors concluded that this issue could be reasonably viewed as a precursor to a significant event and was therefore more than a minor concern. The inspectors also concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective. Specifically, the uncontrolled release of reactor coolant system inventory upset plant stability and challenged the inventory control safety function.
Because Unit 1 was in a shutdown mode during this period, the inspectors performed a Phase 1 SDP review of this issue using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." Based on the above information, the inspectors concluded that the most appropriate Appendix G checklist to use for this issue was the checklist for "Pressurized Water Reactor Hot Shutdown Operation - Time to core boiling less than 2 hours."
Because operator intervention was required to manually close the affected PORV block valve, the inspectors concluded that the unit was in a configuration where a single active failure or personnel error could have resulted in a rapid loss of reactor coolant system inventory as described in Section II.B.(2) of the checklist. Consequently, the inspectors concluded that this issue increased the likelihood of a loss of reactor coolant system inventory and therefore required a Phase 2 SDP analysis. The inspectors discussed the safety significance of this issue with the Regional SRA. The SRA reviewed the finding and determined that the drain path could be easily isolated, accurate reactor coolant system level indication was available, all steam generators were available for cooling, and all trains of standby injection were available and not impacted by the finding. Based on these factors the finding is characterized as having very low safety significance.
Enforcement 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, 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, or drawings. Contrary to the above, the licensee failed to provide a procedure of a type appropriate to the circumstances for testing the Unit 1 pressurizer PORVs, which is an activity affecting quality. Specifically, the instructions contained in 1-IHP-4030-102-017, "Pressurizer Power Operated Relief Valve (PORV) Actuation Channel Calibration with Valve Operation (for Modes 1, 2, and 3)," Revision 1, failed to provide adequate control of pressurizer PORVs 1-NRV-151 and 1-NRV-153, which have a common automatic opening signal. This issue was self-revealed on June 5, 2002, when pressurizer PORV 1-NRV-153 inadvertently opened while testing actuation logic circuitry for 1-NRV-151, causing an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-316-02-09-05(DRP)). The licensee entered this violation into its corrective action program as CR 02157039.
===.2 (Closed) LER 50-315-1999-010-01: "Reactor Coolant System Leak Detection System===
Sensitivity Not in Accordance with TS Basis," Supplement 1. On May 3, 1999, the licensee documented that the Unit 1 and Unit 2 lower containment sump level and flow monitoring capabilities were not consistent with the recommendations of Regulatory Guide 1.45 as stated in TS Basis 3/4.4.6.1. Specifically, the containment sump level and flow monitoring systems were not able to detect a 1 gpm reactor coolant system leak within 1 hour. The licensees review of this issue determined that, other than the TS Basis statement that the containment sump level and flow leak detection systems are consistent with the recommendations of Regulatory Guide 1.45, the most restrictive requirement was that the leak detection system be able to detect a 1 gpm leak within 4 hours to meet Generic Letter 84-04 requirements. The licensee has changed the TS Basis statement using the 10 CFR 50.59 process and documented this corrective action in Supplement 1 to LER 50-315-1999-010-00. The inspectors concluded that this change was appropriate. This event did not constitute a violation of NRC requirements.
This LER is closed.
===.3 Unanticipated Start of the Unit 2 Turbine Driven Auxiliary Feedwater Pump (TDAFWP)===
During a Normal Plant Shutdown for Refueling Outage
====a. Inspection Scope====
On January 19, 2002, in preparation for a Unit 2 refueling outage, operators initiated a planned manual reactor trip of Unit 2 from 22 percent power. Shortly thereafter, an automatic start of the TDAFWP occurred due to a valid low level condition in two of the four steam generators. The inspectors reviewed the circumstances associated with this event, including the root cause determination, operator response leading to and during the event, and corrective actions.
====b. Findings====
(Closed) LER 50-316-2002-004-00: "Unanticipated Start of the Turbine Drive Auxiliary Feedwater Pump."
    (Closed) LER 50-316-2002-004-01: "Unanticipated Start of the Turbine Drive Auxiliary Feedwater Pump," Supplement 1.
    (Closed) LER 50-316-2002-004-02: "Unanticipated Start of the Turbine Drive Auxiliary Feedwater Pump," Supplement 2.
The inspectors identified a finding of very low safety significance (Green) associated with this self-revealed event. The licensee failed to provide instructions of a type appropriate to the circumstances for a planned shutdown of Unit 2, which resulted in unnecessarily challenging the automatic start function of Unit 2 TDAFWP. The inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," and therefore dispositioned this finding as a Non-Cited Violation.
Discussion On January 19, 2002, an automatic start of the Unit 2 TDAFWP occurred due to a valid low level condition in two of the four steam generators following a planned manual reactor trip from 22 percent power in preparation for a refueling outage. The licensee had recently revised its plant shutdown procedure (02-OHP-4021-001-003, "Power Reduction," Revision 15) to trip the reactor from less than 22 percent power in order to enter the refueling outage more expeditiously. The licensee had previously performed the manual reactor trip between 1 percent and 4 percent power. Water levels in each of the four steam generators were within the program band prior to the reactor trip and rapidly shrank to below the TDAFWP auto-start actuation setpoint. Following the reactor trip and TDAFWP start, steam generator water levels rapidly recovered with three AFW pumps supplying water from the condensate storage tank and cooled the reactor coolant system below the system "no-load" temperature of 547 degrees Fahrenheit. Pressurizer level also dropped to below 17 percent as a result of the cooldown, which resulted in reactor coolant system letdown isolation. The automatic start of the TDAFWP and reactor coolant system letdown isolation were both unexpected occurrences for a normal plant shutdown and unnecessarily challenged the operators.
The inspectors previously reviewed the licensees apparent cause evaluation for this event. The NRC concluded in NRC Inspection Report 50-315/316-02-04(DRP) that the licensees ability to consistently identify reasonable causes for conditions adverse to quality in apparent cause evaluations performed for Category 3 CRs was inadequate and documented a finding (FIN 50-315/316-02-04-03). The licensees apparent cause evaluation for this event was one of the four examples included in that finding. The inspectors concluded that the licensees apparent cause evaluation failed to adequately address the cause for the unexpected TDAFWP start. The inspectors noted that the evaluation was limited to the 10 CFR 50.73 reportability aspects of the unexpected actuation of an engineered safety features component. The licensee subsequently wrote CR 02107016 to evaluate the operational aspects of the unexpected automatic pump start and to identify appropriate corrective actions.
The inspectors reviewed the licensees evaluation documented in CR 02107016 and concurred with the licensees conclusion that a planned shutdown should not challenge critical safety equipment to automatically start. The licensee submitted Supplement 1 to LER 50-316-2002-004-00 to provide this conclusion and the corrective actions. The licensee subsequently revised the plant shutdown procedure to initiate the reactor trip from less than 17 percent power and has successfully performed the procedure on both units without challenging the automatic start function of an TDAFWP. The licensee submitted Supplement 2 to LER 50-316-2002-004-00 to identify the cause for the engineered safety features component actuation and to clarify statements made in earlier revisions of the LER. The inspectors concluded that 02-OHP-4021-001-003, "Power Reduction," Revision 15, was not appropriate to the circumstances because initiating the reactor trip from 22 percent power unnecessarily challenged the automatic start function of Unit 2 TDAFWP.
Analysis The inspectors assessed this finding using the SDP. The inspectors concluded that this finding was associated with the initiating events cornerstone and adversely affected the cornerstone objective and was therefore more than a minor concern. Specifically, the function of critical safety equipment was challenged and plant stability was upset during the performance of a normal plant shutdown by the automatic start of Unit 2 TDAFWP.
The inspectors performed a Phase 1 SDP review of this issue using the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because this finding did not cause or contribute to the likelihood of an initiating event, the inspectors concluded that this issue was of very low safety significance.
Enforcement 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, 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, or drawings. Contrary to the above, the licensee failed to provide instructions of a type appropriate to the circumstances for conducting the Unit 2 plant shutdown on January 19, 2002, which is an activity affecting quality. Specifically, the instructions contained in 02-OHP-4021-001-003, "Power Reduction," Revision 15, failed to ensure that the automatic start function of Unit 2 TDAFWP would not be unnecessarily challenged during a normal plant shutdown. Because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-316-02-09-06(DRP)). The licensee entered this violation into its corrective action program as CR 02019036 and CR 02107016.
===.4 (Closed) LER 50-316-2002-005-00: "Unit 2 Trip Due to Instrument Rack 24-Volt DC===
Power Supply Failure." The event described in this LER was discussed in Section 1R12 of this report. The inspectors concluded that the licensees failure to assure that prompt corrective actions were taken to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures was a finding of very low safety significance and a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action". The licensee reported this event as a condition that resulted in an automatic actuation of the reactor protection system in accordance with 10 CFR 50.73(a)(2)(iv)(A). This LER is closed.
===.5 (Closed) LER 50-316-1997-004-02: "Analysis Demonstrates Design Basis Impact of===
Inadequate Refueling Outage Safety Evaluation Was Negligible," Supplement 2. The licensee submitted Supplement 2 to LER 50-316-1997-004-00 to provide additional information concerning the analysis of this event. The inspectors determined that the information provided in Supplement 2 to LER 50-316-1997-004-00 did not raise any new issues or change the conclusion of previous NRC reviews documented in Inspection Reports 50-315/316-97-02(DRP), 50-315/316-98-09(DRS), 50-315/316-99-029(DRS),and 50-315/316-00-01(DRP). This LER is closed.
===.6 (Closed) LER 50-315-1997-005-00: "Reactor Coolant Pump Fire Protection Inoperable===
for Extended Period Without Compensatory Actions Due to Improperly Fabricated Gasket in Spray Header Line." The closure of Supplement 1 to this LER is discussed below in Section 4OA3.7. This LER is closed.
===.7 (Closed) LER 50-315-1997-005-01: "Reactor Coolant Pump Fire Protection Inoperable===
for Extended Period Without Compensatory Actions Due to Improperly Fabricated Gasket in Spray Header Line," Supplement 1. On March 5, 1997, it was discovered that gaskets in the fire protection water spray system for the number 13 reactor coolant pump had not been properly fabricated prior to installation during the 1995 refueling outage. The gaskets for a spectacle flange were fabricated out of a sheet of red rubber, without the center removed to provide a flow path for the water spray. The center area, which should have been cut away during gasket fabrication, was found to be torn. The torn gasket was attributed to the pressure of the system supervisory air. The licensee reported this event in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation prohibited by the plants TSs. Fire Protection requirements were subsequently removed from the TSs in March of 1996. However, at the time the gaskets were installed, the Fire Protection TSs were still in effect. During review of this event, the licensee identified that personnel error was the root cause of the event. The personnel involved did not properly incorporate the information contained in the job order activity regarding fabrication of the gaskets into their actions. At the time of discovery, the licensee corrected the problem and entered the issue into its corrective action program as CR 97-0586. The failure to enter previous TS 3.7.9.2, Table 3.7-5B for an inoperable reactor coolant pump sprinkler system whenever the reactor coolant pump was operable was a violation TS 3.7.9.2. This finding constitutes a violation of minor significance that is not subject to enforcement action in accordance with Section VI of the NRC Enforcement Policy. This LER is closed.
===.8 (Closed) LER 50-315-1998-056-01: "Inadequate Control and Processing of Design===
Information Results in Unanalyzed Hot Leg Recirculation Switchover," Supplement 1.
On December 11, 1998, the licensee identified an unanalyzed condition related to the post-loss-of-coolant accident emergency core cooling system hot leg switchover sub-criticality analysis. The inspectors reviewed the original LER in NRC Inspection Report 50-315/316-99-29(DRS) and concluded that this was a minor issue. The licensee submitted Supplement 1 to LER 50-315-1998-056-00 to provide new information concerning the analysis of the event and corrective actions. The cause of this event was the licensees failure to adequately control design basis calculations and supporting documentation, which is a violation of 10 CFR 50, Appendix B, Criteria III, "Design Control." The licensee entered this event into its corrective action program as CR 98-7848. This finding constitutes a violation of minor significance that is not subject to enforcement action in accordance with Section VI of the NRC Enforcement Policy.
This LER is closed.
===.9 (Closed) LER 50-315-1998-029-01: "Fuel Handling Area Ventilation System Inoperable===
Due to Original Design Deficiency," Supplement 1. The inspectors reviewed the original LER in NRC Inspection Report 50-315/316-99-29(DRS) and concluded that this was a minor issue. The licensee submitted Supplement 1 to LER 50-315-1998-029-00 to provide additional information concerning the analysis of the event, the cause, and the corrective actions. The inspectors determined that the information provided in Supplement 1 to LER 50-315-1998-029-00 did not raise any new issues or change the conclusion of the initial review. This LER is closed.
===.10 (Closed) LER 50-315-1999-003-00: "Control Room Pressurization System Surveillance===
Test Does Not Test System in Normal Operating Condition." On January 7, 1999, the licensee identified that the TS surveillance test procedure for testing the control room pressurization function (12 EHP 4030 STP.229, "Control Room Emergency Ventilation Test," Revision 3,) did not test the control room pressurization system in the normal operating configuration. Specifically, one of the prerequisites prior to performing the surveillance test was to verify that the pressure boundary door that separated the Unit 1 and Unit 2 control rooms (12DR-AUX415) was closed. However, it was identified that the door was normally maintained open to facilitate access and egress between the two control rooms with no procedural guidance to close the door during an event. It was identified that a failure to recognize the door as part of both units control room pressure boundary design resulted in the door being maintained open since initial plant start-up.
The inspectors concluded that this constitutes a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings". The bases for TS Surveillance Requirement 4.7.5.1.e.3 was to demonstrate control room operability such that radiation exposure to personnel occupying the control room would be limited to 5 rem or less whole body, or its equivalent. The licensee as part of its corrective actions closed and labeled pressure boundary door 12DR-AUZ415. In addition, the licensee performed a tracer gas test under several conditions, including one in which door 12DR-AUX415 was left open between the Unit 1 and Unit 2 control rooms with one of the units control room ventilation systems treated as inoperable. The test showed that the amount of unfiltered in-leakage was not highly dependent on pressurization and the dose consequences of having the door open during a postulated accident would remain within 10 CFR 50, Appendix A, General Design Criteria 19 allowable limits. Therefore, this issue is considered to be of minor safety significance. The licensee entered this event into its corrective action program as CR 99-0275. This finding constitutes a violation of minor significance that is not subject to enforcement action in accordance with Section VI of the NRC Enforcement Policy. This LER is closed.
===.11 (Closed) LER 50-315-1999-003-01: "Control Room Pressurization System Surveillance===
Test Does Not Test System in Normal Operating Condition," Supplement 1. The licensee submitted Supplement 1 to LER 50-315-1999-003-00 to provide additional information concerning the analysis of the event, the cause, and the corrective actions.
The inspectors determined that the information provided in Supplement 1 to LER 50-315-1999-003-00 did not raise any new issues or change the conclusion of the initial review which was documented above in Section 4OA3.10. This LER is closed.
===.12 (Closed) LER 50-315-2000-004-00: "Circuit Design Could Result in Failure of===
Emergency Diesel Generators to Load Properly After Loss of Offsite Power." On July 19, 1999, an unanalyzed condition was identified by the Expanded System Readiness Review Teams wherein a sneak electrical circuit existed that could cause improper EDG load sequencing of equipment onto the vital buses following a loss-of-coolant accident concurrent with a loss of offsite power. The licensee reported this event as a condition that was outside the design basis in accordance with 10 CFR 50.73(a)(2)(ii). The licensee implemented a design change in both Unit 1 and Unit 2 to eliminate the possibility of the sneak circuit. This event did not constitute a violation of NRC requirements. This LER is closed.
{{a|4OA5}}
==4OA5 Other==
===.1 (Open) URI 50-316-02-09-07(DRP): "Review of NOED-02-3-058 Regarding D. C. Cook,===
Unit 2, Compliance With Technical Specification 3.8.1.1." By letter dated November 6, 2002, the licensee requested that the NRC exercise discretion not to enforce compliance with the actions of TS 3.8.1.1 regarding operability of the Unit 2 CD EDG. The inspectors opened URI 50-316-02-09-06 to track documentation of the root cause for the Notice of Enforcement Discretion (NOED) request, review the NOED approval basis, and verify licensee activities associated with NOED implementation.
===.2 Completion of Appendix A to Temporary Instruction 2515/148, Revision 1===
The inspectors completed the pre-inspection audit for interim compensatory measures      at nuclear power plants, dated September 13, 2002.
===.3 (Closed) Inspector Follow-up Item (IFI) 50-315/316-99-29-01: "Review and Approval of===
Dose Calculation for General Design Criteria 19 Control Room Habitability Issue." The inspectors reviewed calculation RD-01-05, "Adjusted Dose Consequences for Changes to Control Room," Revision 1. The calculation stated that the control room dose consequences for all events would be below the acceptance criteria required by 10 CFR 50.67. No findings of significance were identified. This item is closed.
===.4 (Closed) IFI 50-316-00-07-03: "Failure to Perform Post Modification Checks to Verify===
Adequate Clearance Between the Pressurizer Surge Line Whip Restraints and the Surge Line Under Hot Plant Conditions." The inspectors performed a limited review of calculation SD-990825-001, "HELB [High Energy Line Break]: Structural Evaluation of Surge Line Pipe Whip Restraints," Revision 3. This calculation included determination of necessary clearances for the pressurizer surge line whip restraints. Additional design engineering documents were reviewed to verify that the calculation results were properly incorporated into the plant design. No findings of significance were identified. This item is closed.
===.5 (Closed) URI 50-315/316-00-16-04: "Determine Whether the Latent Failure of a Test===
Relay Should Be Treated Under the Category of a Single Failure." The NRC staff reviewed this issue and determined that the failure of a K-800 relay would not prohibit the proper operation of an engineered safety features actuation circuit in response to a valid actuation signal. No findings of significance were identified. This item is closed.
===.6 (Closed) URI 50-315/316-01-15-01: "A Change Was Made to the UFSAR Without a===
10 CFR 50.59 Evaluation." The licensee changed the UFSAR and inappropriately used 10 CFR 50.71
: (e) rather than 10 CFR 50.59 to evaluate the UFSAR change. The inspectors reviewed CR 01291058 which described this issue. The licensee corrected this problem by performing a 10 CFR 50.59 screening and concluded that the change did not require NRC approval prior to implementation. No findings of significance were identified. This item is closed.
{{a|4OA6}}
==4OA6 Meetings==
==4OA6 Meetings==
Interim Exit Meeting
* Access Control to Radiologically Significant Areas, Radiation Monitoring Instrumentation and Protective Equipment, and Performance Indicator Verification for RCS Specific Activity with Mr. J. Molden on December 6, 2002.


===.1 Interim Exits===
KEY POINTS OF CONTACT Licensee R. Hershberger, Chemistry Supervisor B. Robinson, Radiation Protection Superintendent M. Scarpello, Regulatory Assurance D. Woods, Radiation/Environmental Manager LIST OF  
 
The results of the Emergency Preparedness Program Inspection were presented to Mr. J. Molden and other members of licensee management at the conclusion of the inspection on December 6, 2002. The licensee acknowledged the findings presented.
 
The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
 
The results of the Radiological Protection Instrumentation and Access Control Inspection were presented to Mr. J. Molden and other members of licensee management at the conclusion of the inspection on December 6, 2002. The licensee acknowledged the findings presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
 
===.2 Resident Inspectors Exit===
 
The inspectors presented the inspection results to Mr. J. Pollock and other members of licensee management at the conclusion of the inspection on January 3, 2003. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
 
No proprietary information was identified.
 
KEY POINTS OF CONTACT Licensee J. Gebbie, Plant Engineering Assistant Director G. Gibson, Site Protective Services C. Graffenius, Emergency Planner S. Greenlee, Nuclear Technical Services Director G. Harland, Work Control/Maintenance Director R. Hershberger, Chemistry Supervisor R. LaBurn, Radiation Protection General Supervisor E. Larson, Operations Director B. McIntyre, Regulatory Assurance Manager R. Meister, Regulatory Affairs Specialist J. Molden, Acting Plant Manager D. Moul, Operation Work Control Manager T. Noonan, Performance Assurance Director S. Partin, Emergency Planning Manager J. Pollock, Site Vice President B. Robinson, Radiation Protection Superintendent M. Scarpello, Regulatory Compliance Supervisor S. Simpson, Operations Staff Manager D. Wood, Radiation / Environmental Manager LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-316-02-09-01      NCV  Failure to assure that prompt corrective actions were taken to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures (Section 1R12)50-316-02-09-02      NCV  Failure to implement a corrective action to prevent recurrence associated with reactor control instrumentation power supply failures (Section 1R12)50-315/316-02-09-03  NCV  Failure to assure that corrective actions were taken to preclude repetition of EDG starting air system relay failures (Section 4OA2.1)50-315-02-09-04      NCV  Failure to identify and take appropriate corrective actions to preclude the failure of reactor coolant system pressure boundary charging line check valves which were at risk of common cause failure due to industry identified design and manufacturing defects (Section 4OA2.2)50-315-02-09-05      NCV  Failure to provide an appropriate procedure for testing the Unit 1 pressurizer power operated relief valves causing an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank (Section 4OA3.1)50-316-02-09-06      NCV  Failure to provide appropriate instructions for a planned shutdown of Unit 2 which resulted in unnecessarily challenging the automatic start function of Unit 2 turbine auxiliary feedwater pump (Section 4OA3.3)50-316-02-09-07      URI  Review of NOED-02-3-058 regarding D. C. Cook, Unit 2, compliance with Technical Specification 3.8.1.1 (Section 4OA5.1)
Closed 50-316-02-09-01      NCV  Failure to assure that prompt corrective actions were taken to address age-related failures of reactor control instrumentation power supplies to prevent repetition of power supply failures (Section 1R12)50-316-02-09-02      NCV  Failure to implement a corrective action to prevent recurrence associated with reactor control instrumentation power supply failures (Section 1R12)50-315/316-02-09-03 NCV Failure to assure that corrective actions were taken to preclude repetition of EDG starting air system relay failures (Section 4OA2.1)50-315-02-09-04    NCV Failure to identify and take appropriate corrective actions to preclude the failure of reactor coolant system pressure boundary charging line check valves which were at risk of common cause failure due to industry identified design and manufacturing defects (Section 4OA2.2)50-315-02-06-01    URI Pressurizer power operated relief valve inadvertently opened during testing resulting in a loss of reactor coolant system inventory and an Unusual Event (Section 4OA3.1)50-315-02-09-05    NCV Failure to provide an appropriate procedure for testing the Unit 1 pressurizer power operated relief valves causing an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank (Section 4OA3.1)50-315-1999-010-01  LER Reactor coolant system leak detection system sensitivity not in accordance with TS [Technical Specification] Basis (Section 4OA3.2)50-316-02-09-06    NCV Failure to provide appropriate instructions for a planned shutdown of Unit 2 which resulted in unnecessarily challenging the automatic start function of Unit 2 turbine auxiliary feedwater pump (Section 4OA3.3)50-316-2002-04-00  LER Unanticipated start of the turbine drive auxiliary feedwater pump (Section 4OA3.3)50-316-2002-04-01  LER Unanticipated start of the turbine drive auxiliary feedwater pump (Section 4OA3.3)50-316-2002-04-02  LER Unanticipated start of the turbine drive auxiliary feedwater pump (Section 4OA3.3)50-316-2002-05-00  LER Unit 2 trip due to instrument rack 24-volt DC [direct current]
power supply failure (Section 4OA3.4)50-316-1997-004-02  LER Analysis demonstrates design basis impact of inadequate refueling outage safety evaluation was negligible (Section 4OA3.5)50-315-1997-005-00  LER Reactor coolant pump fire protection inoperable for extended period without compensatory actions due to improperly fabricated gasket in spray header line (Section 4OA3.6)50-315-1997-005-01  LER Reactor coolant pump fire protection inoperable for extended period without compensatory actions due to improperly fabricated gasket in spray header line (Section 4OA3.7)50-315-1998-056-01  LER Inadequate control and processing of design information results in unanalyzed hot leg recirculation switchover (Section 4OA3.8)50-315-1998-029-01  LER Fuel handling area ventilation system inoperable due to original design deficiency (Section 4OA3.9)50-315-1999-003-00  LER Control room pressurization system surveillance test does not test system in normal operating condition (Section 4OA3.10)50-315-1999-003-01  LER Control room pressurization system surveillance test does not test system in normal operating condition (Section 4OA3.11)50-315-2000-004-00  LER Circuit design could result in failure of emergency diesel generators to load properly after loss of offsite power (Section 4OA3.12)50-315/316-99-29-01 IFI Review and approval of dose calculation for General Design Criteria 19 control room habitability issue (Section 4OA5.3)50-316-00-07-03    IFI Failure to Perform post modification checks to verify adequate clearance between the pressurizer surge line whip restraints and the surge line under hot plant conditions (Section 4OA5.4)50-315/316-00-16-04 URI Determine whether the latent failure of a test relay should be treated under the category of a single failure (Section 4OA5.5)50-315/316-01-15-01 URI A change was made to the UFSAR without a 10 CFR 50.59 evaluation (Section 4OA5.6)
Discussed 50-315/316-02-04-03 FIN Green finding regarding the failure to consistently identify a reasonable apparent cause for conditions adverse to quality (Section 4OA3.1)
LIST OF ACRONYMS USED ADAMS Agency-wide Documents and Management System AFW  Auxiliary Feedwater ALARA As-Low-As-Reasonably-Achievable ANS  Alert and Notification System CDF  Core Damage Frequency CFR  Code of Federal Regulations CO2  Carbon Dioxide CR    Condition Report CY    Calender Year DC    Direct Current DEP  Drill and Exercise Performance DRP  Division of Reactor Projects DRS  Division of Reactor Safety EDG  Emergency Diesel Generator EHP  Electrical Maintenance Head Procedure EP    Emergency Preparedness ERO  Emergency Response Organization ESW  Essential Service Water EWS  Early Warning System FIN  Finding HELB  High Energy Line Break IFI  Inspector Follow-up Item IHP  Instrument Maintenance Head Procedure IMC  Inspection Manual Chapter LER  Licensee Event Report LERF  Larger Early Release Frequency LHRA  Locked High Radiation Area MHP  Maintenance Head Procedure NCV  Non-Cited Violation NEI  Nuclear Energy Institute NOED  Notice of Enforcement Discretion NRC  Nuclear Regulatory Commission NRR  Nuclear Reactor Regulation OA    Other Activities OHP  Operations Head Procedure OWA  Operator Workaround PARS  Publically Available Records PCS  Power Conversion System PI    Performance Indicator PMI  Plant Managers Instruction PMP  Plant Managers Procedure PORV  Power Operated Relief Valve PWR  Pressurized Water Reactor RCS  Reactor Coolant System RPT  Radiation Protection Technician RWP  Radiation Work Permit SDP  Significance Determination Process SPP    Special Plant Procedure SRA    Senior Reactor Analyst SSC    Structures, Systems, and Components STP    Surveillance Test Procedure TDAFWP Turbine Driven Auxiliary Feedwater Pump TS    Technical Specification UFSAR  Updated Final Safety Analysis Report URI    Unresolved Item LIST OF  


=DOCUMENTS REVIEWED=
=DOCUMENTS REVIEWED=


The following is a list of licensee documents reviewed during the inspection, including
The following documents were selected and reviewed by the inspectors to accomplish the
documents prepared by others for the licensee. Inclusion on this list does not imply the NRC
objectives and scope of the inspection and to support any findings.
inspectors reviewed the documents in their entirety, but rather, that selected sections or
20S1
portions of the documents were evaluated as part of the overall inspection effort. Inclusion of a
Access Control to Radiologically Significant Areas
document in this list does not imply NRC acceptance of the document, unless specifically stated
CR 02217009; Modifications to Radiological Posting Program; dated August 5, 2002
in the inspection report.
CR 02226075; Unnecessary Locked High Radiation Areas; dated August 14, 2002
1R01 Adverse Weather Protection
CR 02308023; Valve released to unrestricted area; dated November 4, 2002
PMP 5055-001-001              Winterization/Summerization                Revision 0
CR 02337041; Improper receipt of package containing radioactive source; dated
2-OHP 4022.001.010            Severe Weather                            Revision 1
November 27, 2002
2-IHP 4022.001.009            Plant Winterization and De-Winterization  Revision 0
PMP-6010-RPP-003; High, Locked High, and Very High Radiation Area Access;
CR P-96-00229                  Coils Were Found Frozen and Ruptured      February 14, 1996
Revision 11
on Various Air Handling Units
PMP-6010-RPP-006; Radiation Work Permit Program; Revision 7a
CR P-98-06318                  There is No Freeze Protection for the      October 29, 1998
RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for work at 587'
Condensate Storage Tanks
Drumming room clean-up, dated September 23, 2002
CR P-99-01516                  Plant Does Not Have adequate              January 26, 1999
RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for work on SFP
Winterization Policies
Demineralizer High Pressure Spray of the Inlet Retention Element, dated October 23,
CR P-99-16338                  Winter Storm Damage to Intake              June 10, 1999
2002
Structures
RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for Unit 2 at power
CR P-00-01638                  Unit 1 and 2 Screenhouse Water Level      January 28, 2000
entry to work on No. 1, SI Accumulator; dated November 8, 2002
Sensing Lines Freezing
RWP 020504; Restricted Area NRC Tours and Inspections; Revision 10
CR 02087029                    Reviewed Winterization Program Per        March 28, 2002
RWP 021016; Resin Sluice activities - Locked High Radiation Areas; Revision 6
Condition Report 01318056, Action 1,
RWP 021037; 617 Demin LHRA Work activities; Revision 3
and Determined That Two Procedure
RWP 021046; 587 Drumming Room activities; Revision 1
Enhancements Are Necessary to Make
RWP 021052; Unit 2, At Power Entry; Revision 2
the Program Successful in the Future
RWP 02-1037; Radiation Protection ALARA plan for work on SFP Demineralizer High
1R04 Equipment Alignment
Pressure Spray of the Inlet Retention Element; Revision 1
1R04.1        Partial System Walkdowns
RWP 02-1046; Radiation Protection ALARA plan for work at 587 Drumming room
Unit 2 Turbine Driven and West Auxiliary Feedwater (AFW) System Trains
clean-up; Revision 0
2-OHP-4021-056-001            Filling and Venting Auxiliary Feedwater    Revision 15
RWP 02-1052; Radiation Protection ALARA plan for work on RCP #11 and # 14;
System
Revision 1
2-PMP-4030.001.001            Impact of Safety Related Ventilation in    Revision 5
2-THP-6010-RPP-006; Radiation work Permit (RWP) Processing; Revision 17
the Operability of Technical Specification
2-THP-6010-RPP-401; Performance of Radiation and Contamination Surveys;
Equipment
Revision 10
2-OHP-4030-STP-017W      West Motor Driven Auxiliary Feedwater    Revision 11
2-THP-6010-RPP-418; Radiological Postings; Revision 9
System Test
2OS3 Radiation Monitoring Instrumentation and Protective Equipment
2-OHP-4030-STP-017T      Turbine Driven Auxiliary Feedwater      Revision 15
CR 02246017; Foot and Hand monitor found out of service; dated September 3, 2002
System Test
CR 02249037; Instrument missing from work area; dated September 6, 2002
OP-2-5106A                Flow Diagrams - Auxiliary Feedwater      Revision 45
CR 02287056; Discrepencies between laboratory cross-check program; dated
DB-12-AFWS                Design Basis Document - Auxiliary        Revision 0
Feedwater System
D. C. Cook Nuclear Plant Updated Final  Revision 17
Safety Analysis Report (UFSAR), Section
10.5.2, "Auxiliary Feedwater System"
            (1)
CR 02296004              2-FW-244-2 West Motor Driven AFW        October 23, 2002
Pump Suction Strainer OME-32W South
Basket Vent Valve Leaks at a Rate of
Less than One Drop per Minute
            (1)
CR 02296006              2-FW-244-1 West Motor Driven AFW        October 23, 2002
Pump Suction Strainer OME-32W North
Basket Vent Valve Leaks at a Rate of
Less than One Drop per Minute
            (1)
CR 02298053              2-HV-AFP-FD-4B, Fire Damper for the      October 25, 2002
U2 West Motor Driven Aux Feed Pump
Room, Was Found to Have a Small
Plastic Security Sign Laying in the Fire
Damper Track
Unit 2 AB Emergency Diesel Generator (EDG)
2-OHP-4021-032-008AB    Operating DG2AB Subsystems              Revision 2
OP-2-5251B-59            Flow Diagram Emergency Diesel            Revision 59
Generator AB Unit 2
OP-2-5151A-51            Flow Diagram Emergency Diesel            Revision 51
Generator AB Unit 2
            (1)
CR 02308032              2-DG-136A (2AB EDG Starting Air          November 4, 2002
Receiver Number 2 to Flywheel Air Jack
Connection Shutoff Valve) Was Found
Open During a Procedure Walkdown
Being Performed By an NRC Inspector
Unit 1 West Essential Service Water (ESW) System Train
2-OHP-4021-019-001        Operation of the Essential Service Water Revision 25
System
01-OHP-4030-066-4025        Unit 1 Appendix R and Ventilation        Revision 3
Requirements for Unit 2
OP-1-5113-74                Flow Diagram Essential Service Water    Revision 74
Miscellaneous Condition Reports
              (1)
CR 022700009                Tell Tale Drains From Unit 1 Steam      September 26, 2002
Generator 1 and 4 Safety Valves Have
Signs of Leakage
1R05 Fire Protection
1R05.1        Routine Resident Inspector Tours
: [[contact::D. C. Cook Nuclear Plant UFSAR]],          Revision 17
Section 9.8.1, "Fire Protection System"
D. C. Cook Nuclear Plant Fire Hazards    Revision 8
Analysis, Units 1 and 2
D. C. Cook Nuclear Plant Units 1 and 2  February 1995
Probabilistic Risk Assessment, Fire
Analysis Notebook
D. C. Cook Nuclear Plant Administrative
Technical Requirements Manual,
Sections 1-FP-7 and 2-FP-7, "Fire Rated
Assemblies"
PMP 2270.CCM.001            Control of Combustible Materials        Revision 1
PMP 2270.FIRE.002          Responsibilities for Cook Plant Fire    Revision 0
Protection Program Document Updates
PMP 2270.WBG.001            Welding, Burning and Grinding Activities Revision 0
PMP 5020.RTM.001            Restraint of Transient Material          Revision 1
PMI 2270                    Fire Protection                          Revision 26
2-PPP-2270-066-001        Portable Fire Extinguisher Inspections  Revision 0a
Job Order R0216014          18 Month Surveillance of Fire Dampers in June 9, 2002
Accordance with 12-PPP-4030-066-021
2-PPP-4030-066-021        Inspection of Fire Dampers Protecting  Revision 1
Safety-Related Areas
Job Order R0234423        Perform 6 Month Surveillance of Fire    November 1, 2002
Detection Circuits in Accordance with
2-IHP-4030-STP-206
2-IHP-4030-STP-206        Fire Detection Instrumentation Channel  Revision 3
Functional Test
2-IHP-4030-266-052        Unit 2 Control Rod Drive, Transformer,  Revision 1
Switchgear Room Carbon Dioxide Fire
Suppression Test
Miscellaneous Condition Reports
              (1)
CR 02305079                Misinterpretation of the Administrative November 1, 2002
Technical Requirement Surveillance
Requirements for Fire Damper Closure
Testing
              (1)
CR 02317181                Tracking CR to Add Fire Pump House to  November 13, 2002
Fire Hazard Analysis Fire Zone
Designations
1R11 Licensed Operator Requalification
1R11.1        Resident Inspector Quarterly Review
Licensed Operator Requalification
Training Simulator Evaluation Scenario
for October 29, 2002
1R12 Maintenance Effectiveness
PMP-5035-MRP-001          Maintenance Rule Program                Revision 4
Administration
PMI-5035                  Maintenance Rule Program                Revision 9
Control Group Power Supply Failures
Maintenance Rule (a)(1) Action Plan      Revision 0
Reactor Control 7 Instrumentation
System - Control Group Power Supplies
Maintenance Rule Scoping Document        Revision 1
Reactor Control System
LER 316-2002-005-00      Unit 2 Trip Due to Instrument Rack      July 10, 2002
24-Volt DC (Direct Current) Power Supply
Failure
NRC Information Notice    Inadequate Maintenance of                March 24, 1994
94-24                    Uninterruptible Power Supplies and
Inverters
NRC Information Notice    Potential for Loss of Automatic          August 11, 1995
95-10, Supplement 2      Engineered Safety Features Actuation
NRC Event Notification   
: [[contact::D. C. Cook Unit 2 Tripped From Full      May 12]], 2002
38915                    Power Due to an Instrument Rack Power
Supply Failure
NRC Event Notification   
: [[contact::D. C. Cook Unit 2 Tripped From Full      May 14]], 2002
38915, Revised            Power Due to an Instrument Rack Power
Supply Failure
PMP 4010.TRP.001          Unit Two Reactor Trip Review Report      May 12, 2002
Unit 2 Control Room Logs                May 12, 2002
through
May 13, 2002
CR 01236037              There Have Been a Significant Number    August 24, 2001
of Electronic DC Power Supply Failures
During the Past 24 Months
CR 02047020              2-CG-2-19 Power Supply PS2 Power        February 16, 2002
Available Lamp Is Off
CR 02133001              Both 24-Volt DC Power Supplies in        May 12, 2002
Control Group 1 for Rack 16 Failed
CR 02133002              Unit 2 Trip From 100 Percent Power      May 12, 2002
Level Due to Low Feedwater Flow
Coincident With Low Steam Generator
Level on Loop 1
CR 02133035            After Unit 2 Trip, No Auto-makeup to      May 12, 2002
Volume Control Tank, 2-QRV-303 Went
to Full Divert, and No Refueling Water
Sequence Due to 2-QLC-451
CR 02133058            The Procedure for Volume Control Tank      May 12, 2002
Instrument Malfunction Directs the
Bistable to Be Tripped Which Does Not
Result in a Conservative Condition
            (1)
CR 02134014            TS 3.0.3 Was Entered Erroneously          May 13, 2002
Following the Unit 2 Reactor Trip on May
2, 2002
CR 02137004            This CR Written at Plant Operations        May 17, 2002
Review Committee Chairmans Request
to Drive a Design Engineering Evaluation
of the Instrument Control Power Electrical
Distribution Based on the Equipment
Operability During the Unit 2 Reactor Trip
on May 12, 2002
CR 02138002            24-Volt DC Power Supplies in Control      May 18, 2002
Group 1 Are Only Production 4 Volts DC
CR 02139034            80-Volt DC Power Supply in Control        May 19, 2002
Group 3 1-PS-CGC-20 PS-1 Is Dead
CR 02142030            Unit 2 Tripped Due to Loss of Control      May22, 2002
Room Control Group 1 Power Supplies
            (1)
CR 02325058            Weekly Recurring Tasks to Walkdown        November 21, 2002
Taylor Mod 30 Power Supplies - No
Documented Performance of Walkdown
Since September 30, 2002
CR 02326025            PS2 Available Lamp Not Lit                November 22, 2002
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
PMP-2291-OLR-001        On-Line Risk Management                    Revision 2
PMP-2291-OLR-001        On-Line Risk Management                    Revision 3
NUMARC 93-01            Industry Guideline for Monitoring the      Revision 2
Effectiveness of Maintenance at Nuclear
Power Plants, Section 11, "Assessment
of Risk Resulting From Performance of
Maintenance Activities"
Unit 1 AB EDG
PMP-2291-OLR-001        On-Line Risk Management Work      October 6, 2002
Data Sheet 1            Schedule Review and Approval Form through
Cycle 43, Week 3, with Revisions  October 12, 2002
Unit 2 Control Room Logs          October 8, 2002
through
October 10, 2002
Unit 2 Supervisors Turnover Logs  October 8, 2002
through
October 10, 2002
Unit 2 Abnormal Position Log      October 8, 2002
through
October 10, 2002
Online Integrated Work Schedule  October 8, 2002
through
October 10, 2002
Unit 1 AB EDG ESW Supply Valves
Daily Shift Managers Logs        December 6, 2002
PMP-2291-OLR-001,      On-Line Risk Management Work      December 1, 2002
Data Sheet 1            Schedule Review and Approval Form through
Cycle 43, Week 11                December 7, 2002
Unit 2 CD EDG
Daily Shift Managers Logs        November 4, 2002
through
November 5, 2002
PMP-2291-OLR-001        On-Line Risk Management Work      November 3, 2002
Data Sheet 1            Schedule Review and Approval Form through
Cycle 43, Week 7                  November 9, 2002
Unit 1 East ESW Pump
Daily Shift Managers Logs        December 15, 2002
through
December 19, 2002
PMP-2291-OLR-001        On-Line Risk Management Work      December 15, 2002
Data Sheet 1            Schedule Review and Approval Form through
Cycle 44, Week 1                  December 21, 2002
NRC Letter to
: [[contact::A. Christopher Bakken III]], September 9, 2002
Subject: "Donald
: [[contact::C. Cook Nuclear Plant]],
Units 1 and 2 - Issuance of Amendments
                          (TAC NOS. MB5729 and MB5730)"
Unit 1 West ESW Pump
Daily Shift Managers Logs              October 27, 2002
through
October 31, 2002
PMP-2291-OLR-001          On-Line Risk Management Work            October 27, 2002
Data Sheet 1              Schedule Review and Approval Form        through
Cycle 43, Week 6                        November 2, 2002
Unit 2 East ESW Pump
Daily Shift Managers Logs              November 17, 2002
through
November 23, 2002
PMP-2291-OLR-001          On-Line Risk Management Work            November 17, 2002
Data Sheet 1              Schedule Review and Approval Form        through
Cycle 43, Week 9                        November 23, 2002
NRC Letter to
: [[contact::A. Christopher Bakken III]], September 9, 2002
Subject: "Donald
: [[contact::C. Cook Nuclear Plant]],
Units 1 and 2 - Issuance of Amendments
                          (TAC NOS. MB5729 and MB5730)"
: [[contact::D. C. Cook Nuclear Plant UFSAR]],          Revision 17
Section 9.8.3, "Service Water Systems"
              (1)
CR 02324116              Training for the Sky Jack Fork Truck Was November 20, 2002
Observed Being Performed in Close
Proximity of the Unit 1 Station
Transformer
Unit 2 East AFW System Train
PMP-2291-OLR-001          On-Line Risk Management Work            October 20, 2002
Data Sheet 1              Schedule Review and Approval Form        through
Cycle 43, Week 5, with Revisions        October 26, 2002
Unit 2 Control Room Logs                October 20, 2002
through
October 26, 2002
Unit 2 Supervisors Turnover Logs          October 20, 2002
through
October 26, 2002
Unit 2 Abnormal Position Log              October 20, 2002
through
October 26, 2002
Online Integrated Work Schedule            October 20, 2002
through
October 26, 2002
Miscellaneous Condition Reports
CR 02344008                1-HE-8 Turbine Auxiliary Cooling Water    December 10, 2002
Heat Exchanger Became Air Bound on
the Circulating Water Side of the Heat
Exchanger Resulting in a Secondary
Transient in Unit 1
CR 02346017                There Have Been Repeat Instances of        December 12, 2002
Inconsistencies in Regards to Application
of Cascading TS
CR 02350015                Unit 2 Containment Hydrogen                December 16, 2002
Recombiner Number 2 Maintenance Was
Scheduled During the 1E ESW Pump
Replacement
1R14 Personnel Performance During Non-routine Plant Evolutions
1R14.1        Unit 2 Power Reduction to Support Oil Addition to a Reactor Coolant Pump
Motor
2-OHP-4021-001-003        Power Reduction                            Revision 15
Daily Shift Managers Logs                November 10, 2002
through
November 11, 2002
              (1)
CR 02315078                Rising Pressure Indications on            November 11, 2002
1-IPI-260/-265, Safety Injection Pump
Discharge Pressures
1R14.2        Unit 1 Control Group Power Supply Replacement
PMI-4090                    Infrequently Performed Test or Evolution  December 3, 2002
Briefing Guide for Replacement of PS2
Power Supply at 1-PS-CGC-16
D. C. Cook Nuclear Plant Unit 1
Technical Specifications
Daily Shift Managers Logs                December 3, 2002
1R15 Operability Evaluations
D. C. Cook Nuclear Plant Unit 1 and 2
Technical Specifications
D. C. Cook Nuclear Plant UFSAR            Revision 17
Generic Letter 91-18      Information to Licensees Regarding NRC    Revision 1
Inspection Manual Section on Resolution
of Degraded and Nonconforming
Conditions
PMP-7030-ORP-001          Operability Determinations                Revision 9
R219215-02                Ultrasonic Test Report - Unit 2 Essential December 4, 2002
Service Water Line to West Motor Driven
Auxiliary Feedwater Pump
CR 01101066              Flush Unit 2 ESW Line to West Motor      April 11, 2001
Driven AFW Pump due to Silt/Sand in
Piping
CR 02136008              While Performing Inspection of Termi-    May 15, 2002
point Connection, a Wire Found
Disconnected in 1-RPS-A
CR 02131018              Review Operability and Reportability      May 11, 2002
Issues for Two Items Dealing with
Feedwater Pressure Indication and the
Plant Process Computer Calorimetric
Program
CR 02135049              1-CCR-462 Leaking Excessively During      May 15, 2002
Local Leak Rate Testing
CR 02290012              Steam Generator PORV Actuator            October 17, 2002
Capability Calculation Revealed Negative
Calculated Margin for Full Stroke
Capability
CR 02300002              Unit 2 Control Room Access Door 2-DR-    October 27, 2002
AUX411B Latch Has Broken and Door
Will Not Shut
CR 02339016            Ultrasonic Examination on Unit 1 ESW to  December 5, 2002
West Motor Driven AFW Pump Piping
Found Some Silt/Sand in the Piping
1R16 Operator Workarounds
PMP-4010-OWA-001      Oversight and Control of Operator        Revision 1
Workarounds
NRC Inspection Manual  Evaluation of the Cumulative Effect of
Temporary Instruction  Operator Workarounds
2515/138
Work Around Review Board Meeting        October 23, 2002
Agenda
Work Around Review Board Meeting        October 23, 2002
Minutes
Unit 1 Operator Workarounds              October 23, 2002
Unit 2 Operator Workarounds              October 23, 2002
Unit 1 Operator Workarounds              October 23, 2002
Contingency Actions for Reactor
Operators
Unit 1 Operator Workarounds              October 23, 2002
Compensatory Actions for Auxiliary
Equipment Operators
Unit 2 Operator Workarounds              October 23, 2002
Contingency Actions for Reactor
Operators
Unit 2 Operator Workarounds              October 23, 2002
Compensatory Actions for Auxiliary
Equipment Operators
CR 01048019            Unit 1 Main Turbine Was Deliberately    February 17, 2002
Slowed Due to High Vibration Using the
Vacuum Breakers
CR 01280031            Bypass Steam to Feedwater Heater        October 7, 2001
Valves Leak and Must be Manually
Isolated
CR 02023083            Feedwater Preheating Valves Will Not    January 23, 2002
Isolate Sufficiently to Prevent Cooldown
              (1)
CR 02312059              Update Operations Lesson Plan RQ-C-      November 8, 2002
KNOW. Lesson Plan Contains Incorrect
Information
1R19 Post Maintenance Testing
Unit 2 CD EDG Governor Replacement
2-OHP-4030-STP-        CD Diesel Generator Operability Test    Revision 20
27CD                    (Train A)
PMP-2291-PNT-001        Post Maintenance Testing                Revision 3
2-MHP-5021-032-017      Attachment 1, Guideline To Run Diesel    Revision 4
At Slow Speed
CR 02306005              CD EDG Exhibited 150 Kilowatt            November 2, 2002
Oscillations at Full Load During
Surveillance Testing
Unit 2 East AFW Pump Maintenance
Job Order 02228012-01    2-PP-3E-MTR - Unit 2 East Motor Driven  August 16, 2002
AFW Pump Motor - Check For Soft Foot
and Perform Alignment
Job Order 02136108-01    2-FRV-255 Adjust Packing and Perform    May 16, 2002
Diagnostic Testing
2-OHP-4030-STP-017E    East Motor Driven Auxiliary Feedwater    Revision 10
System Test
2-OHP-4021-056-002,    Auxiliary Feed Pump Operation - East    Revision 13
Motor Driven Auxiliary Feedwater Pump
Long-Term Minimum Flow
2-OHP-4021-056-002,    Auxiliary Feed Pump Operation - East    Revision 13
Motor Driven Auxiliary Feedwater Pump
Operation
Unit 2 East Motor Driven AFW Pump        January 2001
Historical Vibration Data                through
October 2002
OHI-4030                Removal and Restoration of Technical    October 23, 2002
Specification Related Equipment - Unit 2
East Motor Driven AFW Pump
                (1)
CR 02296051                  The Motor Mounting Bolts on the Unit 2    October 23, 2002
East Motor Driven AFW Pump Were Not
Tightened per the Torque Selection
Procedure 12-MHP-5021-001-009
Unit 2 West ESW Pump Maintenance
Job Order 02269031-01        2-PP-7W - Uncouple-Inspect Coupling        September 26, 2002
Gap on Pump
Job Order 02269031-02        2-PP-7W Run Pump/Perform Operability      September 26, 2002
Test
2-OHP-4030-219-022W        West Essential Service Water System        Revision @
Test
Unit 1 Post Accident Containment Hydrogen Monitor
2-EHP-4030-STP-236-        Leak Test of Unit 1 and Unit 2 Post        Revision 1
010                          Accident Containment Hydrogen
Monitoring System
OP-1-5141D-19                Flow Diagram Post-Accident Sampling        Revision 19
Containment Hydrogen Unit No. 1
OP-2-5141D-14               Flow Diagram Post-Accident Sampling        Revision 14
Containment Hydrogen Unit No. 2
1R22 Surveillance Testing
D. C. Cook Nuclear Plant UFSAR            Revision 17
D. C. Cook Nuclear Plant Unit 1 and 2
Technical Specifications
Unit 1 Auxiliary Cable Vault CO2 [Carbon Dioxide] Fire Suppression Test
01-EHP-4030-ATR-225-        Unit 1 Auxiliary Cable Vault CO2 Fire      Revision 0
20                          Suppression Test
Administrative Technical    Section 1-FP-5, Low Pressure CO2          Revision 25
Requirements                Systems
CR 02345019                  Procedure Needs Enhancement to Avoid      December 11, 2002
Unwanted CO2 Discharge
CR 02345020                  Procedure Needs Fixes                      December 11, 2002
Unit 2 Distributed Ignition System Surveillance and Baseline Testing
2-IHP-4030-234-001          Unit 2 Distributed Ignition System        Revision 0
Surveillance and Baseline Testing
: [[contact::D. C. Cook Nuclear Plant UFSAR]],            Revision 17
Section 14.3.6.6, "Distributed Ignition
System"
Job Order R232934-01          Perform Quarterly Distributed Ignition
System Surveillance Unit 2
Steam Generator Steam/Feed Flow Mismatch and Steam Pressure Protection Functional
Testing
2-IHP-4030-SMP-219          Steam Generator 1&2 Steam/Feed Flow        Revision 6
Mismatch and Steam Pressure Protection
Set I Functional Test and Calibration
2-IHP-4030-SMP-222          Steam Generator 2&4 Steam/Feed Flow        Revision 4
Mismatch and Steam Pressure Protection
Set II Functional Test and Calibration
Job Order R0235115-01        Perform 2IHP-4030-SMP-219
PMI-4030 Performance          Performance Review and Acceptance
Review and Acceptance        Sheet for Job Order R0235115-01
Sheet
Job Order R0235114-01        Perform 02-IHP-4030-SMP-222
OP-2-99012                    Steam Generator 1 & 2 Mismatch            Revision 1
Channel 1 Functional Diagram
: [[contact::D. C. Cook Nuclear Plant UFSAR]],            Revision 17
Chapter 7, "Instrumentation and Control"
Containment Isolation and In-service Inspection Valve Operability Testing
01-OHP-4030-STP-011          Containment Isolation and In-service      Revision 23
Inspection Valve Operability Test
Steam Pressure Protection Functional Testing
2-IHP-4030-SMP-227          02-IHP-4030-SMP-227 Steam Pressure        Revision 2
Protection Set III Functional Test and
Calibration
2-IHP-4030-SMP-228        02-IHP-4030-SMP-228 Steam Pressure        Revision 2
Protection Set IV Functional Test and
Calibration
Unit 2 East AFW Pump Characterization Testing
2-EHP-5030-CAR-001        Characterization Testing Program          Revision 0
Motor Analysis Report,      Motor-Driven AFW Pump, 1-PP-3W-MTR        December 4, 2002
Sequence 24
Unit 1 and 2 Personnel Airlock Door Seal Leak Rate Surveillance Testing
2-IHP-4030-046-227        Unit 1 and Unit 2 Personnel Airlock Door  Revision 0
Seal Leak Rate Surveillance
Nuclear Energy Institute    Industry Guideline for Implementing        Revision 0
(NEI) 94-01                Performance-Based Option of 10 CFR
Part 50, Appendix J
Regulatory Guide 1.163      Performance-Based Leak-Test Program        September 1995
NUREG-1493                  Performance-Based Containment Leak-        September 1995
Test Program
Miscellaneous Condition Reports
CR 02269002                Unit 2 Main Turbine "B" Control Valve      September 25, 2002
Opened Unexpectedly From 50 Percent
to 75 Percent, Causing an Unintended
Power Rise and Reactor Coolant System
Temperature Reduction and Automatic
Control Rod Withdrawal
1R23 Temporary Plant Modifications
D. C. Cook Nuclear Plant UFSAR            Revision 17
2-EHP-5040-MOD-001        Temporary Modifications                    Revision 9
Disable East Travel Limit Switch on East Auxiliary Building Crane
2-EHP-5040-EMP-006        Disable Bridge East Travel Limit Switch    Revision 0
on East Auxiliary Building Crane
2-QM-3E
2002-1065-00              10 CFR 50.59 Applicability Determination
for 12-IHP-5040-EMP-006, Revision 0,
                          "Disable Bridge East Travel Limit Switch
on East Auxiliary Building Crane"
2-TC-02-64-R0            Design Packet for Temporary Condition
to Disable East Travel Limit Switch on
East Auxiliary Building Crane
PMP-4050-CHL-001          Control of Heavy Loads                  Revision 1
: [[contact::D. C. Cook Nuclear Plant UFSAR]],          Revision 17
Section 9.7, "Reactor Components and
Fuel Handling System"
: [[contact::D. C. Cook Nuclear Plant UFSAR]],          Revision 17
Section 12.2.1, "Control of Heavy Loads"
Plant Winterization
2-TM-00-61-R2            Winterization/De-Winterization          December 30, 2000
Temporary Modification to Support
2-IHP-5040-EMP-004
Job Order R0235054        Plant Winterization, Perform PM Task 30  September 27, 2002
RPA005058                  Winterization of Tank Vents
2-EHP-5040-MOD-001        Temporary Modifications                  Revision 9
2-IHP-5040-EMP-004        Plant Winterization and De-winterization Revision 3
: [[contact::D. C. Cook Nuclear Plant UFSAR]],          Revision 17
Section 10.5, "Condensate and
Feedwater Systems"
Install Backup Power Supply for Control Group 1
1-TM-02-85-R0              Install Backup Power Supply for Control  November 23, 2002
Group 1
2002-1637-00              10 CFR 50.59 Applicability Determination November 23, 2002
for 1-TM-02-85-R0, "Install Backup
Power Supply for Control Group 1"
Job Order 02326025-01      Install Temporary Modification          November 23, 2002
1-TM-02-85-R0
1EP2 Alert and Notification System (ANS) Testing
Berrien County Early Warning System    December 12, 2001
                          (EWS) Operations Manual
D. C. Cook Sirens and Contour Maps      January 2002
through
September 2002
Berrien County Monthly EWS Test        April 2002
Reports                                through
October 2002
1EP3 Emergency Response Organization (ERO) Augmentation Testing
: [[contact::D. C. Cook Emergency Plan]], Section E    Revision 17
: [[contact::D. C. Cook Emergency Plan]], Section N    Revision 17
PMP-2080-EPP-100            Emergency Response                    Revision 0
PMP-2080-EPP-107            Notification                          Revision 16
SA-2001-SPS-014            Unannounced Drill                      December 14, 2001
SA-2001-SPS-032            Semi-Annual Unannounced Drill          August 23, 2001
SA-2002-SPS-026            Unannounced Drill                      March 14, 2002
SA-2002-SPS-027            Unannounced Drill                      April 16, 2002
SA-2002-SPS-028            On-Shift Unannounced Drill              July 17, 2002
CR 02032031                December 14, 2001 Drill, Emergency      February 1, 2001
Operations Facility and Technical
Support Center Did Not Activate Within
Minutes
1EP5 Correction of Emergency Preparedness Weakness and Deficiencies
PA-01-18                  PA Audit - Emergency Planning          February 25, 2002
PA-02-15                  PA Audit - Emergency Planning          November 22, 2002
PMP-7030-CAP-001            Corrective Action Program Process Flow Revision 13
SA-2001-SPS-026            Self-Assessment - Emergency Plan      July 25, 2001
Graded Exercise
SA-2001-SPS-036 Self-Assessment - 4th Quarter 2001 ERO  December 27, 2001
Drill
SA-2001-SPS-037 Self-Assessment - On-Shift Emergency    February 28, 2002
Planning Staffing Survey
SA-2002-SPS-013 Self-Assessment - 1th Quarter 2002      March 27, 2002
Accountability Drill
SA-2002-SPS-021 Self-Assessment - Emergency Plan Drill  August 7, 2002
SA-2002-SPS-022 Self-Assessment - Emergency Plan Drill  September 24, 2002
SA-2002-SPS-031 Self-Assessment - Review of NRC IN      October 21, 2002
2002-14
CR 01247001    During The ESW Flow Restriction Event    September 3, 2001
on 8/29/01, a More Conservative
Decision Regarding Emergency Plan
Entry Would Have Been Appropriate
CR 02157101    Evaluate Timeliness of NRC Notification  June 6, 2002
For Unusual Event Based On PORV
Opening On June 5, 2002
CR 02163045    Catastrophic Failure Resulting in a Loss June 12, 2002
Of Offsite Power Sources Supplied to
Reserve Feed
CR 02010029    Personnel Manning the New                January 10, 2002
Maintenance Contractor Building Did Not
Report a Failure with the Public Address
System to the Emergency Plan System
Engineer During the Accountability Drill
on January 8, 2002
CR 02204003    ERO Dialogic System Failed to Perform    July 22, 2002
as Required During a Forced Outage
Initiation
CR 02214013    Power Was Lost at the Buchanan Office    August 2, 2002
Building Due to a Storm Resulting in a
Loss of Power to the Emergency
Operations Facility
CR 02268022              The Process for Ensuring Qualified      September 25, 2002
Individuals Report for ERO Duties Failed
and as a Result the Operations Support
Center Was Activated During the 9/18/02
Drill with Unqualified Individuals and
Individuals That Had Not Been Confirmed
to be ERO Qualified
CR 02276062              Error in Minimum Staffing Requirement    October 3, 2002
Found in Table 1, Revision 17 of the
Emergency Plan
CR 02284015              Acceptable Interim Actions Have Not      October 11, 2002
Been Initiated and a Request for Project
Authorization Has Not Been Implemented
to Correct 20 Plant Locations Where it
Has Been Reported That the Public
Address System Is Not in Regulatory
Compliance
CR 02340005              Document Errors Reported in 2nd and 3rd  December 5, 2002
Quarter 2002 NRC DEP Performance
Indicator Data
1EP6 Drill Evaluation
Exercise Scope and Objectives for
October 24, 2002 Annual Exercise
Emergency Notification Forms            October 24, 2002
Completed During Annual Exercise
Desktop Guide for Emergency Planning    Revision 2
Performance Indicators
PMP-2080-EPP-107        Notifications                            Revision 16
2OS1 Access Control to Radiologically Significant Areas
PMP-6010-RPP-003         High, Locked High, and Very High         Revision 11
Radiation Area Access
PMP-6010-RPP-006         Radiation Work Permit Program           Revision 7a
RP 014-01               Total Effective Dose Equivalent         September 23, 2002
Evaluation Worksheet for Work at 587
Foot Drumming Room Clean-up
RP 014-01           Total Effective Dose Equivalent         October 23, 2002
Evaluation Worksheet for Work on Spent
Fuel Pool Demineralizer High Pressure
Spray of the Inlet Retention Element
RP 014-01           Total Effective Dose Equivalent         November 8, 2002
Evaluation Worksheet for Unit 2 at Power
Entry to Work on Number 1, Safety
Injection Accumulator
RWP 020504         Restricted Area NRC Tours and           Revision 10
Inspections
RWP 021016         Resin Sluice activities - Locked High   Revision 6
Radiation Areas
RWP 021037; 617   617 Foot Demineralizer Locked High      Revision 3
Radiation Area Work Activities
RWP 021046; 587   Drumming Room Activities                Revision 1
RWP 021052         Unit 2 At Power Entry                   Revision 2
RWP 02-1037         Radiation Protection ALARA [As-Low-As-  Revision 1
Reasonably-Achievable] Plan for Work
on Spent Fuel Pool Demineralizer High
Pressure Spray of the Inlet Retention
Element
RWP 02-1046         Radiation Protection ALARA Plan for     Revision 0
Work at 587 Foot Drumming Room
Clean-up
RWP 02-1052         Radiation Protection ALARA Plan for     Revision 1
Work on Reactor Coolant Pumps 11 and
2-THP-6010-RPP-006 Radiation Work Permit Processing         Revision 17
2-THP-6010-RPP-401 Performance of Radiation and             Revision 10
Contamination Surveys
2-THP-6010-RPP-418 Radiological Postings                   Revision 9
CR 02217009        Modifications to Radiological Posting    August 5, 2002
Program
CR 02226075        Unnecessary Locked High Radiation        August 14, 2002
Areas
CR 02308023        Valve Released to Unrestricted Area      November 4, 2002
CR 02337041                  Improper Receipt of Package Containing    November 27, 2002
Radioactive Source
2OS3 Radiation Monitoring Instrumentation
PA-02-06                    Performance Assurance Audit, "Radiation    April 16, 2002
Protection"
2-THP06010-RPI-500          Instrument Issue and Operation Testing    Revision 13
2-THP06010-RPI-500          Instrument Issue and Operation Testing;    Revision 13
Data from Portal Monitor Operational
Checks Performed on December 5, 2002
2-THP06010-RPC.512          Calibration of the Eberline Smart Portable Revision 5
Survey Meter(s)
2-THP06010-RPC.512          Calibration of the Eberline Smart Portable Revision 5
Survey Meter(s); Data Sheet from
December 3, 2002
2-THP06010-RPC-513          Calibration of the Eberline Model R0-7    Revision 2
Survey Meter
2-THP06010-RPC-513          Calibration of the Eberline Model R0-7    Revision 2
Survey Meter; Data Sheet from
December 3, 2002
ALARA Radiation Protection Daily Dose      December 2, 2002
Report and Schedule
CR 02246017                 Foot and Hand Monitor Found Out of         September 3, 2002
Service
CR 02249037                 Instrument Missing from Work Area          September 6, 2002
CR 02287056                 Discrepancies Between Laboratory          September 30, 2002
Cross-check Program
CR 02304023                  Failure to Follow Procedural              October 31, 2002
Requirements for Instrument
Accountability
Stations Radiation Protection Instrumentation
Blitz Team Bulletin; Weekly Station        December 3, 2002
Performance Bulletin
Calibration Packages from a Selection of  December 2001
Stations Radiation Protection            through
Instruments                              December 2002
Online Quality Control Schedule          November 27, 2002
Radiation Protection Instrument Use      January 2002
History Analysis Forms; Selections from  through
Number 858-953                            December 2002
Report of Instruments Due for Calibration December 31, 2002
4OA1 Performance Indicator (PI) Verification
NEI 99-02                Regulatory Assessment Performance        Revision 2
Indicator Guideline
SPP-2060-SFI-101        PI Data Gathering                        Revision 0
PMP-7110.PIP.001        Regulatory Oversight Program PI          Revision 1
Letter from
: [[contact::J. Pollock]], American Electric January 17, 2002
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 4Q2001 -- PI Data
Elements (QR and CR)"
Letter from
: [[contact::J. Pollock]], American Electric April 19, 2002
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 1Q2002 -- PI Data
Elements (QR and CR)"
Letter from
: [[contact::J. Pollock]], American Electric July 18, 2002
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 2Q2002 -- PI Data
Elements (QR and CR)"
Letter from
: [[contact::J. Pollock]], American Electric October 21, 2002
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 3Q2002 -- PI Data
Elements (QR and CR)"
Administrative Technical Requirements    Revision 18
Units 1 and 2, Reactor Coolant System,
Supplemental Operational and
Surveillance Requirements
Results of Gamma Spectrometry Count      December 4, 2002
of Units 1 and 2 Reactor Coolant System
Specific Activity Samples
OHI-4032                  Leakage Monitoring Program                Revision 2
2-THP-6020-CHM-101        Reactor Coolant System                    Revision 14
2-EHP-5030-001-008        Recirculation Loop Total Leak Rate        Revision 3
2-THP-6020-INS-026        Gamma Spectrometry System                Revision 1
Licensee Event Reports                    October 1, 2001
through
September 30, 2002
September 30, 2002
NRC Information Notice    Non conservative Reactor Coolant          June 20, 1994
CR 02304023; Failure to follow procedural requirements for instrument accountability;
94-46                      System Leakage Calculation
dated October 31, 2002
CR 01201019                Enhancements May Be Needed in the        July 20, 2001
PA-02-06; Performance Assurance Audit Radiation Protection; dated April 16, 2002
Documentation of Test Results of 4 Alert
2-THP06010-RPI-500; Instrument Issue and Operation Testing; Revision 13a
and Notification System Sirens in the
2-THP06010-RPI-500; Instrument Issue and Operation Testing; Revision 13a; Data
Two State Parks
from Portal Monitor Operational Checks performed on December 5, 2002
CR 01325066                Resident Inspector Observations of Cook  November 21, 2001
2-THP06010-RPC.512; Calibration of the Eberline Smart Portable Survey Meter(s);
Operations Training with Regard to EP
Revision 5
Performance Indicator Data Gathering
2-THP06010-RPC.512; Calibration of the Eberline Smart Portable Survey Meter(s);
CR 02193022                Declining Trend in DEP Emergency          July 12, 2002
Revision 5; Data sheet from December 3, 2002
Planning NRC Performance Indicator
2-THP06010-RPC-513; Calibration of the Eberline Model R0-7 Survey meter;
CR 2009038                Declining Trend in DEP Emergency          January 9, 2002
Revision 2a
Planning NRC Performance Indicator
2-THP06010-RPC-513; Calibration of the Eberline Model R0-7 Survey meter;
CR 02019069                Exceeding Limits for Hard Gammas in      January 19, 2002
Revision 2a; Data sheet from December 3, 2002
Reactor Coolant System Filtrate Isotopic
ALARA Radiation Protection Daily Dose Report and Schedule; dated December 2, 2002
Mixture
Alphabetical listing of stations radiation protection instrumentation
CR 02219004                E-BAR Determinations Found to Be          August 6, 2002
Blitz Team Bulletin; Weekly station performance bulletin; dated December 3, 2002
Slightly Erroneous
Calibration packages from a selection of stations radiation protection instruments; dated
CR P-00-29181              Control Room Operability Evaluation, with December 15, 1999
December 2001 to December 2002
Subsequent Lowering of TS for Reactor
Online Quality Control Schedule; dated November 27, 2002
Coolant System Specific Activity
Radiation Protection Instrument Use History Analysis forms; selections from #858-953;
4OA2 Identification and Resolution of Problems
dated January 2002 to December 2002
4OA2.1        EDG Starting Air Relay Failures
Report of instruments due for calibration; due date of December 31, 2002
JO 00266004                Job Order - Unit 1 CD Diesel Failed to    October 6, 2000
The Plan-It; Daily station performance bulletin; dated December 4, 2002
Stop (Suspect 1-19-DGCD)
4OA1 Performance Indicator Verification
CR P-99-01279          Unit 2 AB EDG Rolled With Air by Itself. January 21, 1999
CR P-00-29181; Control Room Operability Evaluation, with subsequent lowering of
No indication of a Start Signal was
Technical Specification for RCS Specific Activity; dated December 15, 1999 to
Detected Locally or the Control Room.
August 13, 2002
Starting Air Continued to Blow Down
CR 02019069; Exceeding limits for Hard Gammas in RCS Filtrate Isotopic mixture;
Engine Until Air Depleted.
dated January 19, 2002
CR P-99-01336          Found a Relay (EDG - Start Failure Relay  January 22, 1999
CR 02219004; E-BAR determinations found to be slightly erroneous; dated August 6,
2-19-DGAB) in a De-energized State.
2002
Voltage was Present at the Coil
PMP 7110.PIP.001; Regulatory Oversight Program Performance Indicators; Revision 1
Terminals Which Should Have Kept the
PMP 7110.PIP.001; Regulatory Oversight Program Performance Indicators; Revision 1,
Relay in an Energized State
Data Sheet 9, Reactor Coolant system Specific Activity; dated CY 2001, 4th Quarter
CR 00266004            Unit 1 CD Diesel Generator Failed to      September 22, 2000
through CY 2002, 3rd Quarter
Stop from the Control Room
2-THP-6020-INS-026; Gamma Spectrometry System; Revision 1a
CR 02289033            Diesel Generator 2AB 2-OME-150-AB        October 16, 2002
2-THP-6020-CHM-101; Reactor Coolant System; Revision 14c
Starting Rolling Unexpectedly on Starting
Administrative Technical Requirements Units 1 and 2, Reactor Coolant system,
Air
Supplemental Operational and Surveillance Requirements; Revision 18
CR 02296037            Replace 1-19-1-DGC
Results of gamma spectrometry count of Units 1 and 2 RCS Specific Activity samples;
: [[contact::D. There Have          October 23]], 2002
dated December 4, 2002
Been Three 19/19-1 Relay Failures Since
LIST OF ACRONYMS USED
January 1999. They Are Original
ALARA
Installation and Normally Energized.
As Low As Reasonably Achievable
4OA2.2     Common Cause Failure of Four Unit 1 Charging System Check Valves
ARM
PMP 7030.OE.001        Industry Operating Experience            Rev. 5
Area Radiation Monitor
Part 21 Notification, Swing Check Valves  January 18, 1991
CR
                      - Forged Steel
Condition Report
Problem Report 90-1503 Velan Valve Designs                      October 12, 1990
CY
Problem Report 92-157  Part 21 Issue Concerning Velan Valves    February 20, 1992
Calender Year
CR 96-0094            Operating Experience 7640, "Charging      January 24, 1996
LHRA
Injection Valves Found Stuck Open"
Locked High Radiation Area
CR 02132050            Disc on Valve 1-CS-329-L1 Was Found      May 12, 2002
PPT
in the Open Position
Periodic Performance Test
CR 02134021            Check Valves 1-CS-328-L1, 1-CS-328-      May 14, 2002
PI
L4, 1-CS-329-L1, and 1-CS-329-L4 Were
Performance Indicator
Found Open During Radiographic
RCA
Nonintrusive Testing
Radiologically Controlled Area
CR 02205061            10 CFR 21 Evaluation Required for        July 24, 2002
RCS
Manufacturing Defects Discovered on
Reactor Coolant System
Velan 3-inch Check Valves
RP
CR P-00-07039          NRC Information Notice 2000-08            May 16, 2000
Radiation Protection
CR 00278072 Operating Experience Number 11420 October 4, 2000
RPT
CR 01067004 Operating Experience 11950        March 1, 2001
Radiation Protection Technician
CR 01136027 NRC Information Notice 2001-06    May 16, 2001
RWP
CR 01198006 Operating Experience 12454        July 17, 2001
Radiation Work Permit
CR 01198020 Operating Experience 12451        July 17, 2001
SFP
CR 01206021 Operating Experience 12454        July 25, 2001
Spent Fuel Pool
CR 01362008 NRC Information Notice 2001-14    December 28, 2001
SI
CR 02002020 NRC Information Notice 2001-19    January 2, 2002
Safety Injection
4OA3 Event Follow-up
UFSAR
4OA3.1        Pressurizer Power Operated Relief Valve (PORV) Inadvertently Opened
Updated Final Safety Analysis Report
During Testing Resulting in a Loss of Reactor Coolant System Inventory and
an Unusual Event
NRC Event Notification      Unit 1 Declared an Unusual Event Due to    June 6, 2002
38967                        Reactor Coolant System Leakage
Greater Than 25 Gallons-Per-Minute
During Surveillance Testing
01-OHP-4030-102-017          Pressurizer PORV Actuation Channel        Revision 0
Calibration with Valve Operation (for
Modes 1, 2, and 3)
01-OHP-4030-102-017          Pressurizer PORV Actuation Channel        Revision 1
Calibration with Valve Operation (for
Modes 1, 2, and 3)
Daily Shift Managers Logs                June 5, 2002
through
June 6, 2002
CR 02157039                  Pressurizer PORV 1-NRV-153 Opened          June 6, 2002
During Testing with its Block Valve Open,
Causing an Unexpected Release of
Reactor Coolant System Inventory into
the Pressurizer Relief Tank
CR 02157101                  Evaluate Timeliness of NRC Notification    June 6, 2002
for Unusual Event Declared Based on
PORV Opening on June 5, 2002 at 23:00
4OA3.2         LER 315-1999-010-01, "Reactor Coolant System Leak Detection System
Sensitivity Not in Accordance with TS Basis"
LER 315-1999-010-00          Reactor Coolant System Leak Detection      May 3, 1999
System Sensitivity Not in Accordance
with TS Basis
LER 315-1999-010-01          Reactor Coolant System Leak Detection      March 6, 2000
System Sensitivity Not in Accordance
with TS Basis
Generic Letter 84-04        Safety Evaluation of Westinghouse          February 13, 1984
Topical Reports Dealing with Elimination
of Postulated Pipe Breaks in PWR
                            (Pressurized Water Reactors) Primary
Main Loops
4OA3.3      Unanticipated Start of the Unit 2 Turbine Driven Auxiliary Feedwater Pump
            (TDAFWP) During a Normal Plant Shutdown for Refueling Outage
LER 50-316-2002-04-00    Unanticipated Start of the Turbine Drive      March 15, 2002
Auxiliary Feedwater Pump
LER 50-316-2002-04-01    Unanticipated Start of the Turbine Drive      June 28, 2002
Auxiliary Feedwater Pump, Supplement 1
LER 50-316-2002-04-02    Unanticipated Start of the Turbine Drive      December 13, 2002
Auxiliary Feedwater Pump, Supplement 2
NRC Event Notification    The Turbine Driven Auxiliary Feedwater        January 19, 2002
38640                    Pump Auto Started After a Scheduled
Reactor Trip From 20 Percent Power
2-OHP-4021-001-003      Power Reduction                                Revision 15
2001-0985-00              10 CFR 50.59 Screening for Revision 15        January 11, 2002
to 02-OHP-4021-001-003, "Power
Reduction"
Daily Shift Managers Logs                    January 19, 2002
CR 02019036              During Planned Reactor Trip the                January 19, 2002
TDAFWP Started
            (1)
CR 02107016              CR 02019036 Evaluation Did Not                April 17, 2002
Address the Operational Aspects of the
TDAFWP Auto Start on the Planed
Reactor Trip
4OA3.4      LER 50-316-2002-05-00, "Unit 2 Trip Due to Instrument Rack 24-Volt DC
Power Supply Failure"
LER 50-316-2002-05-00    Unit 2 Trip Due to Instrument Rack            July 10, 2002
24-Volt DC Power Supply Failure
4OA3.5      LER 50-316-1997-004-02, "Analysis Demonstrates Design Basis Impact of
Inadequate Refueling Outage Safety Evaluation Was Negligible"
LER 316-1997-04-02        Analysis Demonstrates Design Basis            January 22, 1998
Impact of Inadequate Refueling Outage
Safety Evaluation was Negligible
LER 316-1997-04-01        Change to Component Cooling Water              November 17, 1997
Temperature Without Revision to UFSAR
LER 316-1997-04-00       Change to Component Cooling Water        September 22, 1997
Temperature Without Revision to UFSAR
Results in Condition Outside Design
Basis
CR 97-2342              Inadequate Safety Review Performed for    August 26, 1997
Establishment of a 90EF Upper Limit for
Component Cooling Water During Unit 2
1996 Refueling Outage
Amendment Request        Refueling Operations Decay Time          November 16, 1994
202                    Technical Specification
Amendment Request        Response to Request for Additional        August 1, 1996
202B                    Information (RAI) Technical Specification
Amendment Refueling Operations Decay
Time
Amendment Request        Refueling Operations Decay Time          February 1, 1996
202A                    Updated Analysis and Response to RAI
Amendment Request        Response to RAI Regarding Refueling      June 19, 1997
202D                    Operations Decay Time
Amendment Request        Request to Withdraw the Refueling        January 27, 1998
202F                    Operations Decay Time Technical
Specification Amendment Request
Amendment Request        Refueling Operations Decay Time          November 30, 2001
1146                    Technical Specification
4OA3.6      LER 50-315-1997-005-00, "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly
Fabricated Gasket in Spray Header Line"
LER 315-1997-05-00      Reactor Coolant Pump Fire Protection      April 14, 1997
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
CR 97-0586              When Installing Blank Side of Spectacle  March 5, 1997
Flange Rubber Gasket Not Properly
Installed
4OA3.7        LER 50-315-1997-005-01, "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly
Fabricated Gasket in Spray Header Line"
LER 315-1997-05-00        Reactor Coolant Pump Fire Protection        April 14, 1997
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
LER 315-1997-05-01        Reactor Coolant Pump Fire Protection        October 23, 1997
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
CR 97-0586                When Installing Blank Side of Spectacle    March 5, 1997
Flange Rubber Gasket Not Properly
Installed
4OA3.8        LER 50-315-1998-056-01, "Inadequate Control and Processing of Design
Information Results in Unanalyzed Hot Leg Recirculation Switchover"
LER 50-315-1998-056-00    Inadequate Control and Processing of       January 6, 1999
Design Information Results in
Unanalyzed Hot Leg Recirculation
Switchover
LER 50-315-1998-056-01    Inadequate Control and Processing of        November 24, 1999
Design Information Results in
Unanalyzed Hot Leg Recirculation
Switchover
CR P-98-7848              Unanalyzed Condition Pertaining to Post-    December 11, 1998
Loss-of-Coolant Accident Emergency
Core Cooling System Hot Leg Switchover
4OA3.9        LER 50-315-1998-029-01, "Fuel Handling Area Ventilation System Inoperable
Due to Original Design Deficiency"
LER 50-315-1998-029-01    Fuel Handling Area Ventilation System      August 4, 1999
Inoperable Due to Original Design
Deficiency
Calculation No. RD-99-01  Control Room Dose Resulting from a          Revision 1
Fuel Handling Accident for Off-Load
Specific Conditions
CR P-98-01712            Fuel Handling Area Ventilation System      April 22, 1998
Inoperable Due to Original Design
Deficiency
4OA3.10      LER 50-315-1999-003-00, "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition"
LER 50-315-1999-003-00  Control Room Pressurization System          February 24, 1999
Surveillance Test Does Not Test System
in Normal Operating Condition
CN-CRA-99-78           
: [[contact::D.C. Cook TID-14844 Source Term Loss        February 29]], 2000
of Coolant Accident Radiation Dose
Analysis
EHP 4030 STP 229      Control Room Emergency Ventilation          Revision 3
Test
American Electric Power  NCS Corporation - Control Room              August 11, 1999
Purchase Order A 10342  Envelope In-leakage Testing at D.C.
Cook Nuclear Plant 1999 - Final Report
CR P-99-00275            Control Room Pressurization System          January 7, 1999
Surveillance Test Does not Test System
in Normal Operating Condition
4OA3.11      LER 50-315-1999-003-01, "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition"
LER 50-315-1999-003-01  Control Room Pressurization System          November 10, 2000
Surveillance Test Does Not Test System
in Normal Operating Condition
4OA3.12      LER 50-315-2000-004-00, "Circuit Design Could Result in Failure of
Emergency Diesel Generators to Load Properly After Loss of Offsite Power"
LER 50-315-2000-004-00  Circuit Design Could Result in Failure of  July 3, 2000
Emergency Diesel Generators to Load
Properly After Loss of Offsite Power
NRC Information Notice  Safety Systems Response to Loss of          March 8, 1993
93-17                    Coolant and Loss of Offsite Power
CR P-99-18884            Certain Automatic Safety Systems Could      July 19, 1999
Respond Inappropriately to Certain
Sequences of Loss of Coolant and Loss
of Offsite Power Events
4OA5.3      Inspector Follow-up Item (IFI) 50-315/316-99-29-01, "Review and Approval of
Dose Calculation for General Design Criteria 19 Control Room Habitability
Issue"
DIT-B-00069-00              Design Input for
: [[contact::D. C. Cook Offsite and    July 21]], 1999
Control Room Dose Analysis
DIT-B-00069-09              Design Input for
: [[contact::D. C. Cook Offsite and    April 5]], 2002
Control Room Dose Analysis
RD-01-05                    Adjusted Dose Consequences for            Revision 1
Changes to Control Room
4OA5.4      IFI 50-316-00-07-03, "Failure to Perform Post Modification Checks to Verify
Adequate Clearance Between the Pressurizer Surge Line Whip Restraints and
the Surge Line under Hot Plant Conditions"
2-DCP-4260                  Modification to Surge Line Pipe Whip      Revision 0
Restraints with Field Change Requests
SD-990825-001              HELB Structural Evaluation of Surge Line  Revision 3
Pipe Whip Restraints
CR 01089055                Calculation Impact Assessment Did Not      March 30, 2001
Adequately 03/30/2001 Address Impacts
Associated with Calculation
SD-990825-001
4OA5.5      URI 50-315/316-00-16-04, "Determine Whether the Latent Failure of a Test
Relay Should Be Treated under the Category of a Single Failure"
NRC Task Interface          Evaluation of the Engineering Safety      November 11, 2000
Agreement No. 2000-12      Features Safeguards Test Cabinet
4OA5.6      URI 50-315/316-01-15-01, "A Change Was Made to the UFSAR Without a
CFR 50.59 Evaluation"
CR 01291058                UFSAR Change Request Number 969,          October 18, 2001
Changed the Seismic Class of
Components Within the ESW System,
but Did Not Use CFR50.59 as a Basis for
the Change
(1)
Condition report written as a result of inspection activities.
79
}}
}}

Latest revision as of 12:17, 16 January 2025

Memo 1/10/03, D.C. Cook, DRS Input to DRP IR 05000315-02-009 and 05000316-02-009
ML030630823
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 01/10/2003
From: Slawinski W
Plant Support Branch II
To: Anton Vegel
NRC/RGN-III/DRP/RPB6
References
FOIA/PA-2003-0379, FOIA/PA-2003-0388 IR-01-009
Download: ML030630823 (10)
v • d • e


Text

January 10, 2003 MEMORANDUM TO:

Anton Vegel, Chief Reactor Projects Branch 6 Division of Reactor Projects FROM:

Wayne Slawinski, Acting Chief /RA/

Plant Support Branch Division of Reactor Safety SUBJECT:

D. C. COOK NUCLEAR POWER PLANT DRS INPUT TO INTEGRATED REPORT 50-315/02-09; 50-316/02-09 Attached is the report input for D. C. Cook Nuclear Power Plant, Inspection Report 50-315/02-09; 50-316/02-0 This input provides the results of a recent inspection into aspects of the licensees Occupational Radiation Safety Progra The inspection focused on the licensees Radiological Access Control and Radiation Monitoring Instrumentation and Protective Equipment programs, both under the occupational radiation safety cornerston Also, the performance indicator for the Barrier Integrity cornerstone (i.e., Reactor Coolant System, Specific Activity) was verifie I have reviewed this input and have determined it is ready for distribution to the licensee and dissemination to the public.

After your concurrence, please return this hard copy to me and we will email the electronic version to you.

Attachment:

Input to Inspection Report 50-315/02-09; 50-316/02-09 CONTACT:

Ronald V. Schmitt, DRS (630) 829-9718 DOCUMENT NAME: G:\\DRS\\ML030630823.wpd To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RIII RIII RIII NAME RSchmitt:aa WSlawinski AVegel DATE 1/9/03 1/9/03 1/10/03 OFFICIAL RECORD COPY

Cover Letter



No input, no significant finding Input below, no color or green findings were identified.

Title page Inspector:

Ronald Schmitt, Radiation Specialist

SUMMARY OF FINDINGS

ADAMS boilerplate - Inspectable Area: Occupational Radiation Safety

Modify second paragraph as follows:

The baseline inspection was conducted by a regional radiation specialist inspector.

REPORT DETAILS

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01)

.1 Plant Walkdowns, Radiological Boundary Verifications, and Radiation Work Permit

Reviews

a. Inspection Scope

The inspectors conducted walkdowns of the radiologically protected area to verify the adequacy of radiological area boundaries and postings. Specifically, the inspectors walked down radiologically significant work area boundaries (radiation, high and locked high radiation areas) in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, as well as the Unit 2 Containment. The inspectors performed confirmatory radiation surveys in selected portions of these areas to verify that these areas were properly posted and controlled in accordance with 10 CFR Part 20, licensee procedures, and Technical Specifications. The inspectors also examined the radiological conditions of work areas within those radiation, high and locked high radiation areas to assess the adequacy of licensee implemented contamination controls. Additionally, the inspectors reviewed radiation work permits (RWPs) for general tours, access to locked high radiation areas (LHRAs) for work on spent fuel pool (SFP) demineralizers, drumming room clean-up activities, an at power entry into Unit 1 for work on a reactor coolant pump; and for another at power entry into Unit 2 for work on the no.1 safety injection (SI) accumulator. The RWPs were evaluated for protective clothing requirements, respiratory protection concerns, electronic dosimetry alarm set points, use of remote telemetry dosimetry, radiation protection (RP) hold points, and As-Low-As-Is-Reasonably-Achievable (ALARA) considerations, to verify that work instructions and controls had been adequately specified and that electronic dosimeter set points were in conformity with survey results.

b. Findings

No findings of significance were identified.

.2 Job-In-Progress Reviews, Observations of Radiation Worker Performance, and

Radiation Protection Technician Proficiency

a. Inspection Scope

The inspectors observed selected portions of the following radiologically significant work activities performed during the inspection and evaluated the licensees use of radiological controls:

  • Preparations for SFP demineralizer work The inspectors reviewed the pre-job briefing package for the work evolutions, reviewed the radiological requirements for the activities and assessed the licensees performance with respect to those requirements. The inspectors reviewed survey records, including radiation, contamination, and airborne surveys, to verify that appropriate radiological controls were effectively utilized. The inspectors also reviewed in-process surveys and applicable postings and barricades to verify their accuracy. The inspectors observed radiation protection technician (RPT) and worker performance during the work evolution at the job sites to verify that the technicians and workers were aware of the significance of the radiological conditions in their workplace and RWP controls/limits, and that they were performing adequately given the radiological hazards present and the level of their training.

b. Findings

No findings of significance were identified.

.3 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed licensee Condition Reports (CRs) written since the last inspection (July 2002) to the date of the current inspection, which focused on access control to radiologically significant areas (i.e., problems concerning activities in HRAs, radiation protection technicians performance, and radiation worker practices). The inspectors also reviewed the recently revised High, Locked High, and Very High Radiation Area Access procedure, which addressed new requirements for specific locking devices for these areas. The inspectors reviewed these documents to assess the licensees ability to identify repetitive problems, contributing causes, and the extent of conditions, and then implement corrective actions in order to achieve lasting results.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation (71121.03)

.1 Walkdowns of Radiation Monitoring Instrumentation

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) and performed walkdowns of continuous air monitors in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor, radioactive material building areas, and one area radiation monitor (ARM) in the Unit 2 Containment. Additionally, the inspectors examined a representative number of portable radiation survey instruments staged throughout the licensees facility to verify that those instruments had current calibrations, were operable, and in good physical condition. The inspectors also reviewed the status of repair or troubleshooting activities associated with selected radiation monitoring instruments (i.e. small article monitors and portal monitors that had work request tags)to verify that instrumentation problems were being addressed in an appropriate and timely manner.

b. Findings

No findings of significance were identified.

.2 Calibration, Operability, and Alarm Set Points of Radiation Monitoring Instrumentation

a. Inspection Scope

The inspectors examined radiological instrumentation associated with monitoring transient high and/or very high radiation areas to verify that the instrumentation was operating consistent with industry standards and in accordance with station procedures.

Specifically, the inspectors assessed the operability of the following instrumentation:

  • Unit 2, In-Core Instrumentation Room ARM The inspectors reviewed the licensees alarm set point for this specific ARM to verify that the set point was established consistent with the UFSAR, Technical Specifications, and the stations Emergency Plan.

The inspectors discussed surveillance practices with licensee personnel and reviewed calender year (CY) 2001 - 2002 calibration records and procedures for selected radiation monitors used for assessment of internal exposure. The inspectors also reviewed calibration records and procedures for those instruments utilized for surveys of personnel and equipment prior to egress from the radiologically controlled area (RCA).

These instruments included:

  • AMS-4 Air Monitoring System;
  • APTEC PMW-3 Personnel Monitor; and

Additionally, the alarm set-points for these instruments were reviewed to verify that they were established at levels consistent with industry standards and regulatory guidance provided in Health Physics Positions no. 72 and no. 250 of NUREG/CR-5569.

The inspectors evaluated the calibration procedures and CY 2001 - 2002 calibration records for selected portable radiation survey instruments to verify that the had been properly calibrated consistent with the licensees procedures. Specifically, the inspectors reviewed the calibrations of the following instruments:

  • Smart Radiation Monitor general area dose rate meter.

The inspectors also assessed periodic performance tests (PPTs) completed for selected portable radiation survey instruments to verify that the had been tested consistent with the licensees procedures. Specifically, the inspectors observed the PPTs of the following instruments:

  • Extender instruments; and
  • RSO instruments.

b. Findings

No findings of significance were identified.

.3 Radiation Protection Technician Instrument Use

a. Inspection Scope

The inspectors observed RPTs performing in-field source checks of portable radiation survey instruments to verify that those source checks were adequately completed using appropriate radiation sources and station procedures. The inspectors assessed the RPTs use of radiation/contamination detection instruments as they provided radiological job coverage for risk significant work (e.g. the SI accumulator repair work in the Unit 2 Containment), as well as routine work, to ensure that the RPTs were utilizing the appropriate instruments. The inspectors monitored RPTs performing functional tests of selected contamination monitors, portal monitors, and small article monitors (i.e., for surveys of personnel and equipment prior to unconditional release from the RCA) to verify that they were source tested and calibrated as required by station procedures and industry standards.

b. Findings

No findings of significance were identified.

.4 Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed CY 2001-2002 CRs that addressed radiation monitoring instrument deficiencies to determine if any significant radiological incidents involving instrument deficiencies had occurred. The inspectors examined the results of a self-assessment (i.e., the Summary Report for performance Assurance Audit PA-0206, Radiation Protection) that focused on the licensees CR database and several individual CRs related to radiation monitoring instrumentation generated during the current assessment period. The inspectors also interviewed plant staff and examined closed CRs to verify that radiological instrumentation related issues were adequately addressed by the licensee. The inspectors evaluated these documents to verify the licensees ability to identify repetitive problems, contributing causes, extent of conditions, and the implementation of corrective actions to achieve lasting results.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification

Main article: IP 71151

a. Inspection Scope

The inspectors reviewed the licensees assessment of its performance indicator (PI) for Barrier Integrity, Reactor Coolant System (RCS) Specific Activity. No reportable elements were identified by the licensee for 4th quarter of 2001,and the 1st, 2nd, 3rd quarters of 2002. The inspectors compared the licensees data with CY 2001 and 2002 CRs to verify that there were no occurrences concerning the Barrier Integrity, RCS Specific Activity cornerstone. Additionally, the inspectors also observed staff chemistry technicians collecting RCS samples to verify that the technicians had complied with the applicable procedures during the collection and processing of the samples.

b. Findings

No findings of significance were identified.

4OA6 Meetings

Interim Exit Meeting

  • Access Control to Radiologically Significant Areas, Radiation Monitoring Instrumentation and Protective Equipment, and Performance Indicator Verification for RCS Specific Activity with Mr. J. Molden on December 6, 2002.

KEY POINTS OF CONTACT Licensee R. Hershberger, Chemistry Supervisor B. Robinson, Radiation Protection Superintendent M. Scarpello, Regulatory Assurance D. Woods, Radiation/Environmental Manager LIST OF

DOCUMENTS REVIEWED

The following documents were selected and reviewed by the inspectors to accomplish the

objectives and scope of the inspection and to support any findings.

20S1

Access Control to Radiologically Significant Areas

CR 02217009; Modifications to Radiological Posting Program; dated August 5, 2002

CR 02226075; Unnecessary Locked High Radiation Areas; dated August 14, 2002

CR 02308023; Valve released to unrestricted area; dated November 4, 2002

CR 02337041; Improper receipt of package containing radioactive source; dated

November 27, 2002

PMP-6010-RPP-003; High, Locked High, and Very High Radiation Area Access;

Revision 11

PMP-6010-RPP-006; Radiation Work Permit Program; Revision 7a

RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for work at 587'

Drumming room clean-up, dated September 23, 2002

RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for work on SFP

Demineralizer High Pressure Spray of the Inlet Retention Element, dated October 23,

2002

RP 014-01; Total Effective Dose Equivalent Evaluation worksheet for Unit 2 at power

entry to work on No. 1, SI Accumulator; dated November 8, 2002

RWP 020504; Restricted Area NRC Tours and Inspections; Revision 10

RWP 021016; Resin Sluice activities - Locked High Radiation Areas; Revision 6

RWP 021037; 617 Demin LHRA Work activities; Revision 3

RWP 021046; 587 Drumming Room activities; Revision 1

RWP 021052; Unit 2, At Power Entry; Revision 2

RWP 02-1037; Radiation Protection ALARA plan for work on SFP Demineralizer High

Pressure Spray of the Inlet Retention Element; Revision 1

RWP 02-1046; Radiation Protection ALARA plan for work at 587 Drumming room

clean-up; Revision 0

RWP 02-1052; Radiation Protection ALARA plan for work on RCP #11 and # 14;

Revision 1

2-THP-6010-RPP-006; Radiation work Permit (RWP) Processing; Revision 17

2-THP-6010-RPP-401; Performance of Radiation and Contamination Surveys;

Revision 10

2-THP-6010-RPP-418; Radiological Postings; Revision 9

2OS3 Radiation Monitoring Instrumentation and Protective Equipment

CR 02246017; Foot and Hand monitor found out of service; dated September 3, 2002

CR 02249037; Instrument missing from work area; dated September 6, 2002

CR 02287056; Discrepencies between laboratory cross-check program; dated

September 30, 2002

CR 02304023; Failure to follow procedural requirements for instrument accountability;

dated October 31, 2002

PA-02-06; Performance Assurance Audit Radiation Protection; dated April 16, 2002

2-THP06010-RPI-500; Instrument Issue and Operation Testing; Revision 13a

2-THP06010-RPI-500; Instrument Issue and Operation Testing; Revision 13a; Data

from Portal Monitor Operational Checks performed on December 5, 2002

2-THP06010-RPC.512; Calibration of the Eberline Smart Portable Survey Meter(s);

Revision 5

2-THP06010-RPC.512; Calibration of the Eberline Smart Portable Survey Meter(s);

Revision 5; Data sheet from December 3, 2002

2-THP06010-RPC-513; Calibration of the Eberline Model R0-7 Survey meter;

Revision 2a

2-THP06010-RPC-513; Calibration of the Eberline Model R0-7 Survey meter;

Revision 2a; Data sheet from December 3, 2002

ALARA Radiation Protection Daily Dose Report and Schedule; dated December 2, 2002

Alphabetical listing of stations radiation protection instrumentation

Blitz Team Bulletin; Weekly station performance bulletin; dated December 3, 2002

Calibration packages from a selection of stations radiation protection instruments; dated

December 2001 to December 2002

Online Quality Control Schedule; dated November 27, 2002

Radiation Protection Instrument Use History Analysis forms; selections from #858-953;

dated January 2002 to December 2002

Report of instruments due for calibration; due date of December 31, 2002

The Plan-It; Daily station performance bulletin; dated December 4, 2002

4OA1 Performance Indicator Verification

CR P-00-29181; Control Room Operability Evaluation, with subsequent lowering of

Technical Specification for RCS Specific Activity; dated December 15, 1999 to

August 13, 2002

CR 02019069; Exceeding limits for Hard Gammas in RCS Filtrate Isotopic mixture;

dated January 19, 2002

CR 02219004; E-BAR determinations found to be slightly erroneous; dated August 6,

2002

PMP 7110.PIP.001; Regulatory Oversight Program Performance Indicators; Revision 1

PMP 7110.PIP.001; Regulatory Oversight Program Performance Indicators; Revision 1,

Data Sheet 9, Reactor Coolant system Specific Activity; dated CY 2001, 4th Quarter

through CY 2002, 3rd Quarter

2-THP-6020-INS-026; Gamma Spectrometry System; Revision 1a

2-THP-6020-CHM-101; Reactor Coolant System; Revision 14c

Administrative Technical Requirements Units 1 and 2, Reactor Coolant system,

Supplemental Operational and Surveillance Requirements; Revision 18

Results of gamma spectrometry count of Units 1 and 2 RCS Specific Activity samples;

dated December 4, 2002

LIST OF ACRONYMS USED

ALARA

As Low As Reasonably Achievable

ARM

Area Radiation Monitor

CR

Condition Report

CY

Calender Year

LHRA

Locked High Radiation Area

PPT

Periodic Performance Test

PI

Performance Indicator

RCA

Radiologically Controlled Area

RCS

Reactor Coolant System

RP

Radiation Protection

RPT

Radiation Protection Technician

RWP

Radiation Work Permit

SFP

Spent Fuel Pool

SI

Safety Injection

UFSAR

Updated Final Safety Analysis Report

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000315/2002009&oldid=5426119"