IR 05000293/2005006

IR 05000293-05-006, on 02/14/2005 - 03/03/2005; Pilgrim Nuclear Power Station; Biennial Baseline Inspection of the Identification and Resolution of Problems

ML051050210
Person / Time
Site:	Pilgrim
Issue date:	04/15/2005
From:	Marvin Sykes NRC/RGN-I/DRS/PEB
To:	Balduzzi M Entergy Nuclear Operations
References
IR-05-006
Download: ML051050210 (18)

Text

ril 15, 2005

SUBJECT:

PILGRIM NUCLEAR POWER STATION - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000293/2005006

Dear Mr. Balduzzi:

On March 3, 2005, the NRC completed a team inspection at your Pilgrim Nuclear Power Station. The enclosed inspection report presents the results of that inspection, which were discussed with Mr. Peter Dietrich and your staff on March 3, 2005.

This inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commissions rules and regulations and the conditions of your operating license. Within these areas, the inspection involved examination of selected procedures and representative records, observations of activities, and interviews with personnel.

On the basis of the sample selected for review, the team concluded that in general, problems were properly identified, evaluated, and corrected. The team identified one finding of very low safety significance (Green). This finding was associated with untimely corrective action regarding the February 13, 2005, recurrence of an inoperable condition on the high pressure coolant injection (HPCI) system due to a faulty fuse, similar to a February 2004 incident. This finding was determined to be a violation of NRC requirements. However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating this finding as a non-cited violation, in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny this non-cited violation, you should provide a response with the basis for your denial within 30 days of the date of this inspection report to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at the Pilgrim Nuclear Power Station.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publically Available Records (PARS) component of the NRCs 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).

Mr. Michael Balduzzi 2 If you have any questions, please contact me at (610) 337-5046.

Sincerely,

/RA by Joseph G. Schoppy Acting for/

Marvin D. Sykes, Chief Performance Evaluation Branch Division of Reactor Safety Docket No. 50-293 License No. DPR-35

Enclosure:

Inspection Report 05000293/2005006 w/Attachment: Supplemental Information

REGION I==

Docket No: 50-293 License No: DPR-35 Report No: 05000293/2005006 Licensee: Entergy Nuclear Operations, Inc.

Facility: Pilgrim Nuclear Power Station Location: 600 Rocky Hill Road Plymouth, MA 02360 Dates: February 14 - March 3, 2005 Inspectors: G. Meyer, DRS, Senior Reactor Inspector (Team Leader)

M. Davis, DRS, Reactor Inspector B. Sienel, DRP, Resident Inspector (Vermont Yankee)

J. Talieri, DRS, Reactor Inspector Approved by: Marvin D. Sykes, Chief Performance Evaluation Branch Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS IR 05000293/2005006; 02/14/2005 - 03/03/2005; Pilgrim Nuclear Power Station; biennial baseline inspection of the identification and resolution of problems; problem identification and resolution.

This team inspection was performed by three regional inspectors and a resident inspector from another site. One finding of very low safety significance (Green) was identified during this inspection and was classified as a non-cited violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. 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.

Identification and Resolution of Problems The team determined that Entergy was generally effective at identifying problems and placing them in the corrective action program (CAP). Once entered into the system, these items were screened and prioritized in a timely manner using established criteria, and they were properly evaluated commensurate with their safety significance. Overall, the evaluations reasonably identified the causes of the problem, assessed the extent of condition, and developed appropriate corrective actions. Corrective actions were typically implemented in a timely manner, but the team found that in one case, corrective actions were not timely and did not prevent recurrence; this resulted in a finding. The team found that Entergys self-assessments and audits were self-critical and consistent with the teams observations.

A. NRC Identified and Self-Revealing Findings Cornerstone: Mitigating Systems

• Green. Entergys corrective actions were untimely, in that faulty Bussmann fuses which had caused an inoperable condition on HPCI in February 2004 were not replaced and caused another unplanned HPCI inoperable condition in February 2005. Entergy did not take timely action to establish the extent of affected fuses, determine priorities for replacement, and replace the faulty fuses. The team determined that this represented a self-revealing non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI of very low safety significance (Green). This finding was associated with the cross-cutting area of problem identification and resolution (PI&R).

ii Enclosure

This finding was more than minor, because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of assuring the availability and reliability of systems that respond to initiating events. The finding was determined to be of very low safety significance based on a Phase 3 SDP determination. (Section 4OA2.c)

B. Licensee-Identified Violations None.

iii Enclosure

Report Details 4. OTHER ACTIVITIES (OA)

4OA2 Problem Identification and Resolution a. Effectiveness of Problem Identification (1) Inspection Scope The team reviewed the procedures describing the CAP at Entergys Pilgrim Nuclear Power Station. Entergy identifies problems by initiating condition reports (CRs) for conditions adverse to quality, plant equipment deficiencies, industrial or radiological safety concerns, or other significant issues. Condition reports are subsequently screened for operability, categorized by significance level (A through D), and assigned to personnel for evaluation and resolution.

The team reviewed items selected across the seven cornerstones of safety in the NRCs Reactor Oversight Program to determine if problems were being properly identified, characterized, and entered into the CAP for evaluation and resolution. The team selected items from the maintenance, operations, engineering, emergency planning, security, radiological control, and oversight programs to ensure that Entergy was appropriately considering problems identified in each. The team considered risk insights from Entergys probabilistic risk assessment (PRA) to focus the sample selection and system walkdowns on risk-significant components. The team used this information to select a risk-informed sample of CRs that had been issued since the last NRC PI&R inspection, which was completed in January 2003.

The team reviewed a sample of Entergys audits and self-assessments, including an audit of the CAP. This review was performed to determine if problems identified through these evaluations were entered into the CAP, and whether the corrective actions were properly completed to resolve the deficiencies. The effectiveness of the audits and self-assessments was evaluated by comparing audit and self-assessment results against self-revealing and NRC-identified findings.

Based on a review of NRC and Entergy risk analyses, the team selected five high risk-significance systems (residual heat removal (RHR), reactor building closed cooling water (RBCCW), 4 kVAC, 480 VAC, and HPCI) to focus the review of corrective action processes. For the selected risk-significant systems, the team reviewed a sample of applicable system health reports, work requests, engineering documents, plant log entries, and results from surveillance tests and maintenance tasks. For these selected systems, the team also interviewed cognizant station personnel and completed system walkdowns to assess material condition and system performance.

In addition, the team interviewed plant staff and management to determine their understanding of and involvement with the CAP. The specific documents reviewed and referenced during the inspection are listed in the attachment to this report.

Enclosure

(2) Observations and Findings No findings of significance were identified.

The team concluded that the station was generally effective at problem identification.

Entergy staff generally had adequate knowledge of the CAP, and identified problems and entered them into the program at an appropriate threshold. There were few deficiencies identified by the team that had not been previously identified by Entergy.

Station staff promptly initiated CRs, as appropriate, in response to deficiencies or issues raised by the inspection team.

The team found that self-assessments and audits were self-critical and generally consistent with the teams observations, and that identified issues were appropriately addressed in the CAP.

b. Prioritization and Evaluation of Issues (1) Inspection Scope The team reviewed the CRs listed in the attachment to this report to assess whether Entergy adequately prioritized and evaluated problems. The team selected the CRs in areas to cover the seven cornerstones of safety in the NRCs Reactor Oversight Program. The team also considered risk insights from Entergys PRA to focus the inspection sample in general with emphasis on the five selected risk-significant systems.

The reviews included the appropriateness of the assigned significance level, the timeliness of problem resolution, and the scope and depth of the causal analysis. For significant conditions adverse to quality, the team reviewed Entergys assessment of the extent of condition and the determination of corrective actions to preclude recurrence. A portion of the items chosen for review was expanded to five years. The team observed Condition Review Group (CRG) meetings, in which Entergy managers reviewed incoming CRs and evaluated preliminary corrective action assignments, analyses, and plans.

In addition, the team selected a sample of CRs written to address previous NRC NCVs to determine whether Entergy evaluated and resolved these problems. The team reviewed Entergys evaluation of industry operating experience information for applicability to Pilgrim. For applicable CRs, the team reviewed Entergys assessment of equipment operability and reportability requirements.

Enclosure

(2) Observations and Findings No findings of significance were identified.

The team concluded that Entergy generally screened and evaluated problems at the correct significance level. The staff was generally effective at classifying and performing operability evaluations and reportability determinations for discrepant conditions.

Additionally, the team determined that the CRG was effective in reviewing and prioritizing CRs, and evaluating causal analyses at a plant management level.

The team reviewed several root cause evaluations and found that they were generally adequate. In most cases, the evaluations were thorough and corrective actions would be reasonably expected to prevent recurrence.

c. Effectiveness of Corrective Actions (1) Inspection Scope The team reviewed the corrective actions associated with selected CRs to determine whether the actions addressed the identified causes of the problems. The team reviewed CRs for repetitive problems to determine whether previous corrective actions were effective. The team also reviewed Entergys timeliness in implementing corrective actions and their effectiveness in precluding recurrence of significant conditions adverse to quality. Furthermore, the team assessed the backlog of corrective actions to determine if any, individually or collectively, represented an increased risk due to delays in implementation. The team also reviewed NCVs issued since the last inspection of Entergys CAP to determine if issues placed in the program had been properly evaluated and corrected. The team also attended the February 15 and 17 Corrective Action Review Board meetings.

(2) Observations and Findings Overall, the team concluded that Entergy developed and implemented corrective actions that were appropriate and effective. Based on the sample reviewed, the team determined that corrective actions were completed in a timely manner. Nonetheless, the team determined that in one instance corrective actions for a previous event did not prevent recurrence, because they were not effectively implemented or timely.

Introduction. Entergys corrective actions were untimely, in that faulty fuses which had caused an inoperable condition on HPCI in February 2004 were not replaced and caused another HPCI inoperable condition in February 2005. The team determined that this represented a self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI of very low safety significance (Green). Specifically, in response to a February 2004 spurious failure of a Bussmann fuse, Entergy did not take timely action to establish the extent of affected fuses, determine priorities for replacement, and replace the faulty Bussmann fuses. As a result, when another Bussmann fuse failed in the control power circuit for the HPCI injection valve, the HPCI system was inoperable on February 13, 2005.

Enclosure

Description. On February 26, 2004, a control power fuse on the HPCI gland seal condensate pump spuriously blew and made the HPCI system inoperable until the fuse was replaced. Entergy found that the fuse had lost continuity due to the detachment of a cold-solder connection between the fusible link and the end cap. NRC determined that Entergy had performed an ineffective review of previous industry-wide information on such fuse failures and issued a Green NCV (NCV 05000293/2004004-001) for the failure to identify and replace the faulty fuses.

Following the 2004 fuse failure, Entergy began addressing the susceptible Bussmann fuses still installed in the plant, including identifying the location of the susceptible fuses, classifying each identified fuse according to safety significance and whether or not the fuse failure would be apparent, and planning the fuse replacements systematically according to that classification scheme. Entergy changed relevant maintenance procedures to ensure examination and replacement of susceptible fuses would be performed during scheduled system outage windows.

At the time of the fuse failure on the HPCI injection valve in February 2005, Entergy had identified all susceptible fuses, but had not yet completed the classification, and had not begun a systematic replacement of the fuses. Entergy had completed the procedure changes to replace fuses, but without the systematic approach in place, there was no way to ensure safety-related systems would receive prompt attention to correct the problem.

After the February 2005 event, Entergy completed the classification process, issued work orders for the highest priority fuses, and began to schedule fuse replacement work for the upcoming spring outage.

Analysis. The performance deficiency is that Entergy did not take timely corrective actions after the February 2004 HPCI event identified a condition adverse to quality. As a result, HPCI experienced another inoperable condition in February 2005 due to an identical failure in a Bussmann fuse - this time in the control power circuit for the HPCI injection valve.

The finding is more than minor, because it is associated with the equipment performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective. Entergy did not take prompt action to classify and replace faulty Bussmann fuses which resulted in the HPCI system being inoperable on February 13, 2005.

Enclosure

The team evaluated the issue using the SDP Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones provided in IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding affected the Mitigating Systems cornerstone and resulted in the loss of the safety function of the HPCI system. Therefore, this finding required a Phase 2 evaluation.

In accordance with IMC 0609, Appendix A, Attachment 1, User Guidance for Significance Determination of Reactor Inspection Findings for At-Power Situations, the senior reactor analyst evaluated the finding using the Risk-Informed Inspection Notebook for Pilgrim Station, Revision 1. The analyst made the following assumptions:

• The HPCI system was not functional upon failure of the injection valve control power fuse and the would not have responded upon demand.

• While the fuse was failed, the HPCI system could not have been recovered prior to postulated core damage because of the following:

- No direct indication existed that the fuse had failed (only that a control power failure had occurred).

- Test equipment would have been required to determine that the fuse had failed.

- A replacement fuse would not have been immediately available to the operators.

• The HPCI system was unavailable for a maximum of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This was considered the maximum time the system could have been unavailable without the control room operators knowledge. This includes one operating shift plus operator turnover time. Consequently, the Phase 2 exposure time used to select the initiating event likelihood in Table 1 of the Risk-Informed Inspection Notebook was < 3 days.

For a finding involving degradation of the HPCI system, Table 2 of the Risk-Informed Inspection Notebook requires all of the Phase 2 SDP worksheets be evaluated except for large a loss of coolant accident and loss of the B 125 vdc bus. All core damage sequences involving HPCI system operation were evaluated. The most significant core damage sequence involved a transient with loss of the power conversion system followed by failure of the operators to manually depressurize the reactor with safety relief valves. Using the counting rule worksheet, this finding was estimated to be White for internal initiators.

Given the that the finding was potentially greater than Green in significance, the analyst performed a Phase 3 SDP analysis for internal initiators using the Standardized Plant Analysis Risk (SPAR) model for Pilgrim Station, Revision 3.11. The same assumptions were used as described above for the Phase 2 analysis, with the exception that an actual HPCI exposure time of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> was used in the SPAR model simulation. The Enclosure

analyst performed a condition assessment using the Pilgrim Station SPAR model with Basic Event HCI-MOV-CC-HPCI, HPCI Injection Valve Fails to Open, set to TRUE.

This adequately modeled the as found condition.

When evaluated over the exposure time of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, the incremental conditional core damage probability (the change in core damage frequency over the exposure time)

associated with this finding was 2.4E-9. Similar to the Phase 2 SDP Worksheet result, the most significant core damage sequence involved a loss of main feedwater event with loss of the power conversion system and operator failure to manually depressurize the reactor.

The analyst concluded the SPAR model result was a reasonable estimation of the risk associated with this finding. The Phase 2 SDP Notebook result was conservative because the initiating event likelihood used was based on exposure time of 0.01 year (87.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) and the SPAR model analysis used the actual maximum exposure time of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. The analyst also determined that an evaluation of risk resulting from external initiators or risk from a large early release frequency perspective was not required. This was because the result of the Phase 3 internal event significance estimation was less than 1E-7. Therefore, this was a finding of very low safety significance (Green).

This finding is cross cutting in the area of PI&R, because Entergy did not take prompt action to correct a significant condition adverse to quality.

Enforcement. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that significant conditions adverse to quality be promptly identified and corrected. Entergy procedure ENN-LI-102, Corrective Action Process specifies that conditions adverse to quality be reviewed and evaluated, and that corrective actions be taken to preclude repetition. Contrary to the above, following the identification that failed control power Bussmann fuses had affected components in the HPCI system in February 2004 and in earlier instances, Entergy did not take prompt action to correct the problem and HPCI was again inoperable in February 2005.

Because the finding is of very low safety significance and has been entered into Entergys CAP (CR 200500517), this violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy. (NCV 0500293/2005006-001, Untimely Corrective Action for Bussmann Fuses)

d. Assessment of Safety Conscious Work Environment (1) Inspection Scope The team interviewed various plant personnel to develop a general perspective of the safety-conscious work environment (SCWE), including whether employees were reluctant to raise safety concerns. Additionally, the team reviewed Entergys Employee Concerns Program (ECP) to evaluate if employees were aware of the program and had used it to raise concerns.

Enclosure

(2) Observations and Findings No findings of significance were identified.

The team determined that individuals were aware of the importance of nuclear safety, stated a willingness to raise safety issues, had not experienced retaliation in any prior issues raised, and had an adequate knowledge of the CAP and ECP. Based on these limited interviews, the team concluded that there was no evidence of an unacceptable SCWE. Also, the team noted that the ECP had demonstrated effective involvement in raising and addressing concerns, including some regarding significant operational and support activities.

4OA6 Meetings, including Exit The team presented the inspection results to Mr. Peter Dietrich and members of the Entergy staff on March 3, 2005. No proprietary information was retained by the team.

ATTACHMENT: SUPPLEMENTAL INFORMATION Enclosure

A-1 ATTACHMENT SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee Personnel S. Beneduci I&C Superintendent D. Burke Security Supervisor G. Choquette RBCCW System Engineer W. Corbo Mechanical Superintendent D. Detterman Chemistry Supervisor P. Dietrich General Manager, Pilgrim F. Famulari Mechanical Maintenance Corrective Actions Coordinator B. Ford Licensing Manager S. Hudson Maintenance Rule Coordinator D. Landeche Corrective Action and Assessment Manager W. Lobo Licensing Engineer F. Marcussen Security Manager J. Martin Electrical System Engineer F. Mulcahy HPCI System Engineer D. Noyes Asst. Operations Manager - Operations Support E. Olson Operations Manager D. Perry Radiation Protection Manager D. Rydman RHR System Engineer N. Santiago Employee Concerns Coordinator T. Sowdon Emergency Planning Superintendent T. Trask System Engineering Manager J. Tucker FIN Team Supervisor LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened and Closed 05000293/2005006-01 NCV Untimely Corrective Action for Bussmann Fuses. (Section 4OA2.c)

Attachment

A-2 LIST OF DOCUMENTS REVIEWED Procedures EN-LI-102 Corrective Action Process, Revision 1 EN-LI-111 Operational Decision-making Issue (ODMI) Process, Revision 2 EN-LI-118 Root Cause Analysis Process, Revision 0 EN-OE-100 Operating Experience Program, Revision 1 EN-WM-100 Work Request (WR) Generation, Screening, and Classifications, Revision 0 ENN-DC-121 Maintenance Rule, Revision 2 ENN-OP-104 Operability Determinations, Revision 2 1.3.34.4 Compensatory Measures, Rev. 1 1.3.121.3 Supplemental Guidance for Implementing the PNPS Corrective Action Program, Rev. 1 1.4.5 Tagging Procedure, Rev. 66 2.1.1 Startup from Shutdown, Rev. 137 2.1.4 Approach to Critical, Rev. 20 2.1.26 Inventory of Alternate Shutdown and EOP Support Tools and Materials, Rev. 25 2.1.41 Torus Water Cleanup by Processing Through Radwaste, Rev. 4 2.2.19 Residual Heat Removal, Rev. 84 2.2.19.1 RHR System - Shutdown Cooling Mode of Operation, Rev. 11 2.2.21 High Pressure Coolant Injection System, Rev. 65 2.2.87.3 Control Rod Drive Venting, Timing, and Adjustment, Rev. 11 2.2.87.3 Control Rod Drive Venting, Timing, Flushing, and Adjustment, Rev. 19 2.2.90 Rod Worth Minimizer, Rev. 22 5.3.35 Operations Management Emergency and Transient Response Expectations for Operating Crews, Rev. 6 8.5.2.2.2 LPCI System Loop B Operability - Pump Quarterly and Biennial (Comprehensive) Flow Rate Tests and Valve Tests, Rev. 28 8.5.3.1 RBCCW System Quarterly Operability, Rev. 39 8.5.3.18 RBCCW System Biennial Comprehensive Operability, Rev. 5 8.5.4.1 HPCI System Pump and Valve Quarterly Operability, Revs. 72 and 75 8.5.4.3 HPCI Operability Demonstration and Flow Rate Test at 150 PSIG, Rev. 39 8.5.4.12 Manual Start of the HPCI Turbine for Maintenance Activities, Rev. 6 8.5.5.1 RCIC Pump Quarterly and Biennial Operability Flow Rate and Valve Test at Approximately 1000 PSIG, Rev. 55 3.M.1-34 Generic Troubleshooting and Maintenance Procedure, Revision 25 3.M.4-79 HPCI Turbine Preventive Maintenance Inspection, Revision 6 3.M.4-81 HPCI Stop Valve Balance Chamber Adjustment, Revision 8 3.M.4-84 HPCI turbine Mechanical-Hydraulic Overspeed Trip Inspection, Revision 4 8.M.2-2.10.3-3RHR Shutdown Cooling Valve Interlock Test, Rev. 7 8.M.2-2.10.5 HPCI Auto-Isolation System Logic, Rev. 22 8.5.4.1 HPCI System Pump and Valve Quarterly Operability, Revision 75 8.E.23 HPCI System Instrumentation Calibration, Revision 47 8.Q.3-3 480V AC Motor Control Center Testing and Maintenance, Revision 42.

Attachment

A-3 Audits and Self-Assessments QA-03-2004-PNP-01 Corrective Action Program QA-02-2004-PNP-01 Chemistry QA-07-2004-PNP-01 Emergency Preparedness Program QA-03-08 AC Power System PNP-LO-2003-37 Management Ownership of Corrective Actions PNP-LO-2003-44 Emergency Preparedness NRC Performance Indicators PNP-LO-2003-50 Periodic Assessment of the Maintenance Rule PNP-LO-2003-63 Human Performance Tool Usage in Maintenance PNP-LO-2003-85 Emergent Work / Work Prioritization Process PNP-LO-2003-87 Temporary Alteration Process PNP-LO-2003-92 Radiation Protection - Human Performance PNP-LO-2003-96 Radiation Protection PNP-LO-2004-07 Operations Attention to Detail PNP-LO-2004-20 Radiological Surveys and Documentation PNP-LO-2004-33 Effectiveness of Corrective Actions for Operations Training AFIs PNP-LO-2004-36 Mechanical Maintenance Training Programs PNP-LO-2004-45 QA Organization Effectiveness PNP-LO-2004-61 Security Work Hour Controls Pilgrim Cross-Functional Corporate Assessment Report, March 18, 2004 Pilgrim Operations Self-Assessment Report, August 8, 2003 Pilgrim Forced Outage Assessment, April 15, 2003 Condition Reports (CR-PNP-XXX, unless noted)

2002-0222 2003-1517 2003-3258 2004-0047 2004-1049 2002-10792 2003-1663 2003-3302 2004-0168 2004-1107 2002-10824 2003-1695 2003-3304 2004-0189 2004-1138 2002-11406 2003-1788 2003-3305 2004-0358 2004-1270 2002-11421 2003-1790 2003-3321 2004-0374 2004-1301 2002-12035 2003-2137 2003-3324 2004-0420 2004-1303 2002-12251 2003-2145 2003-3507 2004-0422 2004-1488 2002-12550 2003-2159 2003-3530 2004-0501 2004-1489 2002-12573 2003-2356 2003-3546 2004-0623 2004-1619 2002-13003 2003-2716 2003-3594 2004-0624 2004-1670 2003-0071 2003-2721 2003-3597 2004-0705 2004-1684 2003-0113 2003-2729 2003-3601 2004-0706 2004-1754 2003-0133 2003-2792 2003-3827 2004-0733 2004-1907 2003-0398 2003-2805 2003-3831 2004-0742 2004-1941 2003-0651 2003-2859 2003-3957 2004-0781 2004-2158 2003-0698 2003-2860 2003-4008 2004-0799 2004-2198 2003-0735 2003-2895 2003-4159 2004-0812 2004-2279 2003-0737 2003-2931 2003-4366 2004-0818 2004-2327 2003-0940 2003-3035 2003-4387 2004-0821 2004-2377 2003-1012 2003-3044 2003-4424 2004-0868 2004-2395 2003-1507 2003-3045 2003-4493 2004-0980 2004-2397 Attachment

A-4 2004-2429 2004-2973 2004-3744 2005-0065 2005-0347 2004-2497 2004-3131 2004-3752 2005-0105 2005-0354 2004-2571 2004-3137 2004-3820 2005-0149 2005-0435 2004-2722 2004-3245 2004-3832 2005-0183 2005-0517 2004-2788 2004-3265 2004-3834 2005-0235 2005-0526 2004-2792 2004-3495 2004-3868 2005-0240 2005-0582 2004-2862 2004-3505 2004-3916 2005-0243 2005-0643 2004-2918 2004-3595 2004-3952 2005-0321 2005-0655 2004-2971 2004-3708 2005-0059 2005-0322 NRC Non-Cited Violations0500293/2003003-001 0500293/2003006-002 0500293/2004005-001 0500293/2003004-001 0500293/2003006-003 0500293/2004008-001 0500293/2003004-002 0500293/2003011-001 0500293/2004008-002 0500293/2003005-001 0500293/2003011-003 0500293/2003006-001 0500293/2004004-001 Operating Experience Documents IN 1987-62 Mechanical Failure of Indicating-Type Fuses IN 1999-14 Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, and FitzPatrick IN 2000-01 Operational Issues Identified in Boiling Water Reactor Trip and Transient IN 2002-05 Foreign Material in Standby Liquid Control Storage Tanks IN 2002-14 Ensuring a Capability to Evacuate Individuals, Including Members of the Public, From the Owner-Controlled Area IN 2004-05 Spent Fuel Pool Leakage to Onsite Groundwater Maintenance Requests 04100608 04101805 04114728 04101801 04102939 Engineering Requests 04113073 04113316 04118395 System Health Reports 4Q 2004 HPCI 4Q 2004 RHR 4Q 2004 RBCCW Miscellaneous Attachment

A-5 Operations Standards Operations Human Performance Action Plan Maintenance Rule SSC Basis Document - RBCCW System (30a), Rev. 1 Simulator Discrepancy Report DRA3-020, A RHR min flow valve logic ODMI Implementation Plan, Fuel Defect, January 18, 2005 LIST OF ACRONYMS ADAMS Agencywide Document Management System CAP Corrective Action Program CFR Code of Federal Regulations CR Condition Report CRG Condition Review Group ECP Employee Concerns Program HPCI High Pressure Coolant Injection IMC Inspection Manual Chapter IN Information Notice NCV Non-Cited Violation NRC Nuclear Regulatory Commission PARS Publically Available Records PI&R Problem Identification and Resolution PRA Probabilistic Risk Assessment RBCCW Reactor Building Closed Cooling Water RHR Residual Heat Removal SCWE Safety-Conscious Work Environment SDP Significance Determination Process SPAR Standardized Plant Analysis Risk Attachment