IR 05000333/2004006
| ML042570100 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/10/2004 |
| From: | Ray Lorson NRC/RGN-I/DRS/PEB |
| To: | Ted Sullivan Entergy Nuclear Northeast |
| References | |
| IR-04-006 | |
| Download: ML042570100 (15) | |
Text
September 10, 2004
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC PROBLEM IDENTIFICATION & RESOLUTION INSPECTION REPORT 05000333/2004006
Dear Mr. Sullivan:
On August 5, 2004, the NRC completed an inspection at the James A. FitzPatrick Nuclear Power Plant. The enclosed report documents the inspection findings which were discussed on August 5, 2004, with you and members of your staff.
The inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, 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 samples selected for review, there were no findings of significance identified during this inspection. The team concluded that problems were properly identified, evaluated, and resolved within the problem identification and resolution program. However, during the inspection, some examples of minor problems were identified related to long-standing and recurring equipment deficiencies that were not effectively evaluated and corrected in a timely fashion.
In accordance with 10 CFR 2.390 of the NRC's "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 Publicly 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).
Sincerely,
/RA by C.G. Cahill Acting for/
Raymond K. Lorson, Chief Performance Evaluation Branch Division of Reactor Safety Docket No.
50-333 License No.
Mr. Theodore Sullivan
Enclosure:
Inspection Report No. 05000333/2004006 w/Attachment: Supplemental Information
REGION I==
Docket No:
50-333 License No:
DPR-59 Report No:
05000333/2004006 Licensee:
Entergy Nuclear Northeast (Entergy)
Facility:
James A. FitzPatrick Nuclear Power Plant Location:
268 Lake Road Scriba, New York 13093 Dates:
July 19, 2004 - August 5, 2004 Inspectors:
Stephen M. Pindale, Senior Reactor Inspector (Team Leader)
Brice A. Bickett, Reactor Inspector Douglas A. Dempsey, Resident Inspector (FitzPatrick)
Brian J. Fuller, Resident Inspector (Nine Mile Point)
June Cai, Human Performance Analyst (Observer)
Approved by:
Raymond K. Lorson, Chief Performance Evaluation Branch Division of Reactor Safety
Enclosure ii SUMMARY OF FINDINGS IR 05000333/2004006; 7/19/04 - 7/23/04 and 8/2/04 - 8/5/04; James A. FitzPatrick Nuclear Power Plant; biennial baseline inspection of the identification and resolution of problems.
This inspection was conducted by two regional inspectors and two resident inspectors. 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 NRC team determined that Entergy was effective at identifying discrepant conditions at an appropriate threshold and entering them into the corrective action program. Once entered into the system, issues were typically prioritized appropriately and in a timely fashion; and were properly evaluated commensurate with the safety significance. Overall, the evaluations reasonably identified the causes of the problem, the extent of the condition, and provided for corrective actions to address the causes. However, the team noted some minor instances where long-standing and recurring equipment problems were not effectively evaluated and corrected in a timely fashion. On the basis of interviews conducted, the team determined that plant staff personnel were familiar with and utilized the corrective action program to identify problems.
A.
NRC-Identified and Self-Revealing Findings No findings of significance were identified.
B.
Licensee-Identified Violations None.
Enclosure Report Details 4.
OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution a.
Effectiveness of Problem Identification (1)
Inspection Scope The inspection team reviewed the procedures describing the corrective action program (CAP) at the James A. FitzPatrick Nuclear Power Plant. The team reviewed items selected from various Entergy processes and activities to determine whether personnel were properly identifying, characterizing and entering problems in the CAP for evaluation and resolution. Entergys formal CAP utilizes condition reports (CRs) to identify and document problems at FitzPatrick. The team reviewed a sample of CRs as well as maintenance work orders to cover the seven cornerstones of safety identified in the NRC Reactor Oversight Process (ROP). In addition, the team considered risk insights from the individual plant examination report and the probabilistic risk assessment to focus the sample selection and system walkdowns on risk significant components. The CRs are classified by category level (A, B, C, D) with level A requiring the most rigorous review due to higher safety and/or risk significance.
The team reviewed logs, control room deficiencies, operator work-arounds, system health reports, temporary modifications, operating experience reviews, and procedures.
The team selected items from Entergys maintenance, operations, engineering, emergency planning, security, radiological controls and oversight processes for entry into the CAP. In addition, the team interviewed plant staff and management to determine their understanding of and involvement with the CAP; and to determine whether personnel were familiar with and utilized the CAP to identify problems. The specific documents reviewed and referenced during the inspection are listed in the attachment to this report.
The team reviewed a sample of quality assurance audits and surveillances, and departmental self-assessments. The review was to determine whether the problems identified by these assessments were entered into the CAP, and whether the corrective actions were properly completed to resolve the self-identified deficiencies. The team evaluated the effectiveness of the audits and self-assessments by comparing the associated results against self-revealing and NRC-identified findings.
The team also conducted several plant walkdowns of safety-related, risk significant areas to determine if observable system equipment and plant material adverse conditions were identified and entered into the CAP. Team members attended daily review and management meetings where CRs were reviewed for screening and assignment. The team attended these meetings to understand the threshold for identifying problems and to assess management involvement with the CAP. The team also assessed the interface between the CAP and the work control process.
Enclosure (2)
Observations and Findings No findings of significance were identified.
The team identified only minor deficiencies where CRs had not been previously initiated; and for those identified by the team, Entergy promptly initiated CRs to address the deficiencies. Accordingly, the team concluded that plant staff identified deficiencies and entered them in the CAP, and at an appropriate threshold. The team also found that self-assessments and audits were sufficiently self-critical and provided relevant performance observations and insights.
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. These reviews evaluated the causal assessment of each issue (i.e., root cause analysis, apparent cause evaluation);
and for significant conditions adverse to quality, the extent of condition and determination of corrective actions to preclude recurrence. The team selected the CRs to cover the seven cornerstones of safety identified in the NRC ROP. A portion of the items chosen for review were those that were age dependent (e.g., service water system erosion and/or corrosion, heat exchanger fouling), and accordingly, the scope of review was expanded to five years. The team also considered risk insights from the FitzPatrick probabilistic risk assessment to help focus the inspection sample.
Throughout the inspection, the team attended periodic meetings to observe the CR review process and to understand the bases for assigned category and root cause level.
The team selected a sample of CRs associated with previous NRC non-cited violations (NCVs) and findings to determine whether Entergy evaluated and resolved problems associated with compliance with applicable regulatory requirements and standards. The team reviewed Entergys evaluation of industry operating experience for applicability to FitzPatrick. The team also reviewed Entergys assessment of equipment operability and reportability requirements associated with CRs.
(2)
Observations and Findings No findings of significance were identified.
Overall, the team found that CRs were appropriately prioritized and evaluated. The quality and completeness of root cause evaluations and apparent cause analyses were generally good. In addition, the team observed that both the CR Screening Committee and the CR Review Group were effective in reviewing and prioritizing CRs. The team identified some minor instances where the bases or justification for specific actions associated with CRs were not well documented or readily apparent. Examples included the following:
Enclosure
CR 2003-05720 (maintenance preventable functional failure determination associated with the reactor vessel level instrumentation system);
CR 2003-1333 (emergency service water system weld pinhole leak/extent of condition review);
DER 2002-4980 (corrective actions did not get into CAP system associated with a RCIC test valve stroke issue); and
Several CRs (reactor core isolation cooling system/piping inspection not performed/documented).
Station personnel subsequently addressed each of these minor documentation deficiencies.
c.
Effectiveness of Corrective Actions (1)
Inspection Scope The team reviewed the corrective actions associated with selected CRs to determine whether the actions had addressed the identified cause(s) of the problems. The team also reviewed Entergys timeliness for implementing the corrective actions, and their effectiveness in precluding recurrence for significant conditions adverse to quality.
Additionally, the team assessed the backlog of outstanding corrective actions to determine if they, individually or collectively, represented an increased risk to the plant.
The team also reviewed the NCVs and findings issued since the last inspection of the FitzPatrick CAP to determine if issues placed in their program had been properly evaluated and corrected.
(2)
Observations and Findings No findings of significance were identified.
The team determined that overall, corrective actions associated with CRs were appropriate and effective. However, there were some instances where long-standing and recurring equipment problems were not effectively evaluated and/or corrected.
Three specific examples were identified, which were related to 1) excessive seat leakage from the high pressure coolant injection (HPCI) system steam supply admission valve 23MOV-14; 2) degraded performance of a safety-related cooler for electrical switchgear (67UC-16B); and 3) repeated instances of pressurizing the reactor core isolation cooling (RCIC) system discharge piping slightly higher than its design value during quarterly tests. Details for each of these examples are provided below.
- HPCI steam supply admission valve (23MOV-14) has had a long-standing history (about five years) of excessive seat leakage, which has resulted in several adverse effects on the HPCI system (e.g., minor corrosion/pitting of the turbine rotor, water intrusion in the turbine lube oil). Entergy had identified that past maintenance practices were less than adequate, and subsequently overhauled the valve in February 2003 to correct the problem. While the leakage initially abated, the valve subsequently began leaking again, and is currently exceeding
Enclosure design seat leakage. Entergys actions in February 2003 were ineffective in achieving lasting success with regard to preventing excessive seat leakage. In response to the continuing leakage, Entergy has implemented a monitoring program with appropriate action if HPCI turbine temperatures indicate a degrading trend (CR-JAF-2000-04897 tracks this issue).
- Safety-related cooler for electrical switchgear 67UC-16B was degraded such that the cooler may not have been capable of removing postulated accident heat loads if the ultimate heat sink temperature was at its maximum assumed value of 85F. This condition has existed for the past four years. Actions taken to date to ensure the cooler can consistently satisfy the thermal performance test have been untimely and ineffective. Further, the degraded condition remained a challenge to the organization because additional testing and analysis are required each time the cooler fails to satisfy the thermal performance test.
Based upon existing analysis and actual temperature conditions, the cooler remains operable. Entergy plans to replace this cooler during the next refueling outage (CR-JAF-2004-01519 and CR-JAF-2004-02688).
- RCIC over-pressurization during quarterly surveillance. Four events (in 1997, 1998, 2002, 2003) occurred where discharge piping was pressurized slightly beyond its design pressure of 1320 psig. While the surveillance procedure limit is 1235 psig, it appears the cause is that the test return valve, 13MOV-30, is overly sensitive to adjustments in valve position. Therefore, small adjustments in valve position results in relatively large pressure changes. Previously, plant staff had chosen to live with this issue and rely on engineering analysis to allow over-pressure up to 1400 psig (prior surveillance procedure changes were not designed to correct the adverse condition). Entergy was pursuing possible design and/or procedure changes to prevent further similar challenges (CR-JAF-2003-1182).
The team reviewed the details associated with each of the three items above, and determined that equipment operability has not been adversely impacted; and in each case, Entergy was pursuing an appropriate corrective action plan.
Enclosure 4OA6 Meetings, including Exit The team presented the inspection results to M and other members of Entergy management and staff on August 5, 2004. Entergy acknowledged that no proprietary information was involved.
A-1 Attachment ATTACHMENT SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee Personnel S. Bono, Manager, System Engineering J. Boyer, System Engineering Manager G. Brownell, Regulatory Compliance B. Burnham, System Engineer A. Degracia, System Engineer D. Denbleyker, Employee Concerns Coordinator T. Edwards, System Engineer J. Fischer, System Engineer A. Halliday, Manager, Regulatory Compliance S. Haskell, System Engineer D. Huwe, QA Auditor D. Johnson, Manager, Operations T. Johnson, Maintenance Supervisor M. Kayhan, System Engineer W. Maguire, Director, Nuclear Safety K. Mulligan, General Manager, Plant Operations D. Nacamuli, Self-Assessment Coordinator J. Pechacek, Manger, Engineering Support W. Rheaume, CA&A Manager D. Ruddy, Design Engineering Supervisor L. Stoner, Auxiliary Operator T. Sullivan, Vice President, Operations R. Thomas, Control Room Supervisor D. Wallace, Quality Assurance Manager K. Wells, Senior Nuclear Operator LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED NONE
A-2 Attachment LIST OF DOCUMENTS REVIEWED Procedures AP-12.11 Response to Operational Concerns and Notifications, Rev. 2 [[::JAF-LI-102|JAF-LI-102]] JAF Corrective Action Process, Rev. 1 ENN-LI-102 Corrective Action Process, Rev. 4 ENN-OP-104 Operability Determinations, Rev. 2 ENN-LI-104 Self-Assessment and Benchmark Process, Rev. 4 EN-LI-104 Self-Assessment and Benchmark Process, Rev. 0 ENN-WM-101 On-Line Work Management Process, Rev. 0 ODSO-17 Operator Plant Tour and Operating Logs, Rev. 74 OP-22 Diesel Generator Emergency Power, Rev. 49 ST-24J RCIC Flow Rate and Inservice Test, Rev. 32 Condition Reports (all are prefixed with CR-JAF-)
1997-00019 1999-02296 1999-02700 1999-03037 2000-00385 2000-02386 2000-04897 2001-00783 2001-03848 2001-04683 2002-00415 2002-02215 2002-02520 2002-02713 2002-02720 2002-02721 2002-03081 2002-03132 2002-03232 2002-03279 2002-03552 2002-03602 2002-03667 2002-03716 2002-04006 2002-04008 2002-04303 2002-04351 2002-04742 2002-04773 2002-04794 2002-04946 2002-04986 2002-04993 2002-05148 2002-05295 2002-05505 2003-00177 2003-00188 2003-00305 2003-00307 2003-00309 2003-00526 2003-00733 2003-00794 2003-00854 2003-00863 2003-00918 2003-00968 2003-01088 2003-01182 2003-01333 2003-01418 2003-01581 2003-01843 2003-01844 2003-02172 2003-02251 2003-02327 2003-02432 2003-02528 2003-02639 2003-02827 2003-02847 2003-02861 2003-03024 2003-03028 2003-03162 2003-03456 2003-03592 2003-03921 2003-04049 2003-04362 2003-04382 2003-04556 2003-04566 2003-04585 2003-04625 2003-04906 2003-05560 2003-05613 2003-05720 2003-05899 2004-00224 2004-00289 2004-00411 2004-00618 2004-00685 2004-00699 2004-00720 2004-00721 2004-00725 2004-00908 2004-00971 2004-01167 2004-01256 2004-01519 2004-01573 2004-01947 2004-02023 *
2004-02425 2004-02651 2004-02760 2004-02875 2004-02876 2004-02877 2004-02878 2004-02960 2004-02979 2004-03033 *
2004-03034 2004-03151 *
2004-03157 *
2004-03160 *
2004-03172 *
2004-03173 *
(Note * = CR was generated as a result of NRC inspection)
A-3 Attachment Operating Experience CRs CR-OEN-2002-00080 CR-OEN-2002-00109 CR-OEN-2002-00205 CR-OEN-2002-00220 LO-OEN-2002-00232 CR-OEN-2003-00008 LO-OEN-2002-00034 LO-OEN-2002-00072 LO-OEN-2002-00198 Audits and Self-Assessments A03-07J Technical Specifications A03-09J Results of Actions to Correct Deficiencies (2003 Audit)
JAFLO-2003-00007 Condition Reports Closed by Corrective Action Coordinators JAFLO-2003-00010 Benchmarking Trip to Perry JAFLO-2003-00013 Work Planning and Preventive Maintenance Feedback JAFLO-2003-00022 Control room activities JAFLO-2003-00034 Communications JAFLO-2003-00051 Operability Determination Process JAFLO-2003-00062 Snapshot Assessment on System Health Report Process JAFLO-2003-00076 Surveillance testing JAFLO-2003-00095 Effectiveness Review of 2001 Audit/Assessment Corrective Actions JAFLO-2003-00096 Security Post Instructions JAFLO-2003-00097 ALARA/Radiation Dose Feedback JAFLO-2003-00129 Operations Training JAFLO-2003-00148 Snapshot Assessment: Closure of Corrective Actions to Other Processes JAFLO-2003-00241 Security Firing Range JAFLO-2003-00256 Conduct of Maintenance QS-2004-JAF-005 Operations QA Findings QS-2004-JAF-002 Testing of Hand Geometry System QA-4-2004-JAF-1 Design Control (LO-JAF-2004-00005)
SR 2303 Surveillance Report - Review of Actions to Correct Deficiencies SR-2325 Oversight of Engineering and 10CFR50.59 Activities SR-2330 Abnormal Plant Conditions/Long term Unexplained Indications SR-2343 ISFSI Corrective Actions SR 2358 Blackout Post-transient Review and Recovery Plan
A-4 Attachment Maintenance Rule (a)(1) Action Plans and Basis Documents:
Automatic Depressurization System (safety/relief valve pilot valve leakage)
RHR check valves Decay Heat Removal (flow rate)
NSW/ESW systems Uninterruptible Power Supply motor-generator set System Health Reports (First Quarter 2004)
DC Electrical Distribution System Reactor Water Recirculation Nuclear Boiler Instrumentation Reactor Protection System Residual Heat Removal System (and 4th Qtr 2003, 2nd Qtr 2004)
Reactor Core Isolation Cooling System High Pressure Coolant Injection System Emergency Service Water System (and 4th Qtr 2003)
Normal Service Water System (and 4th Qtr 2003)
Work Orders [[::JAF-03-27088|JAF-03-27088]] JAF-03-27088 [[::JAF-03-33777|JAF-03-33777]] JAF-04-10672 [[::JAF-04-12427|JAF-04-12427]] JAF-04-13633 [[::JAF-04-18338|JAF-04-18338]] JAF-04-24392 [[::JAF-04-24436|JAF-04-24436]] JF-011411100 JF-020767600 JF-020767600 JF-980142904 Engineering Requests [[::JAF-04-13211|JAF-04-13211]] JAF-04-17580 JF-03-00003 JF-03-00091 JF-03-00589 JF-03-00909 JF-03-01056 JF-03-01105 JF-03-01305 JF-03-01450 JF-03-01656 JF-03-01858
A-5 Attachment Other Performance Engineering Oil Analysis Data - RHRSW 2002-2004 JENG-APL-01-001, HPCI Improvement Action Plan, Rev. 2 Calculation JAF-CALC-HPCI-02133, Thrust and Torque Limits Calculation for 23MOV-19" Root Cause Evaluation Report - Reactor Pressure Vessel Level Transient Coincident with B Station Battery Ground Fault During HPCI Surveillance Test, June 22, 2004 LIST OF ACRONYMS ADAMS Agencywide Documents Access & Management System CA&A Corrective Action and Assessment CAP Corrective Action Program CFR Code of Federal Regulations CR Condition Report DC Direct Current ESW Emergency Service Water HPCI High Pressure Coolant Injection ISFSI Independent Spent Fuel Storage Installation NCV Non-Cited Violation NRC Nuclear Regulatory Commission NSW Normal Service Water PARS Publically Available Records psig Pounds per Square Inch - Gauge QA Quality Assurance RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water ROP Reactor Oversight Process