IR 05000354/1997005
| ML20198H745 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/15/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198H743 | List: |
| References | |
| 50-354-97-05, 50-354-97-5, NUDOCS 9709220158 | |
| Download: ML20198H745 (43) | |
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{{#Wiki_filter:. Y U. S. NUCLEAR REGULATORY COMMISSION
REGION I
D'-ck A No: 50-354 Li'en e Nos: NPF-57 Report N /97-05 Licensee: Public Service Electric and Gas Company Facility: Hope Creek Nuclear Generating Station Location: P.O. Box 236 Hancocks Bridge, New Jersey 08038 Dates: July 13,1997 - August 23,1997 Inspectors: S. A. Morris, Senior Resident Inspector J. D. Orr, Reactor Engineer Approved by: James C. Linville, Chief, Projects Branch 3 Division of Reactor Projects l l l i j 9709220158 970915 , PDR ADOCK 05000354 l G PDR l t
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EXECUTIVE SUMMARY Hope Creek Generating Station NRC Inspection Report 50 354/97-05 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspectio Operations in spite of frequent challenges from both unexpected and planned even, the operations department exhibiteJ improved performance during the period. Communications, command and control, and coupliance with TS requirements were all good. Based on a review of the monthly QA report and discussion with various staff, the inspectors judged that oversight of operations department activities was appropriat Maintenance Maintenance technicians appropriately implemented troubleshooting activities following unexpected failures of important to safety equipment, however the quality of the troubleshooting plans varied. Supervision at job sites was evident, however in one case both a supervisor and manager failed to identify missed steps in a preventive maintenance procedure. Good use of operating experience, both internal and external, was observe Monthly emergency diesel generator tests were well coordinated, planned and execute Flaws in procedure quality were noted, both in pre-start check activities and test implementation methodology, Identified deficiencies, including a proceduralized preconditioning issue, indicated weaknesses in attention-to-detail during procedure development and usag The Quality Assurance department performed frequent and appropriate evaluations of maintenance activities. Responses to Filtration, Recirculation, and Ventilation System surveillance testing issues were reasonable, timely, and effectiv Enaineerina t l Implementation of PSE&G's maintenance rule program for the station service water system l was generally good in that the system performance was appropriately classified, monitored ! and evaluated. Weaknesses were evident in individual monitoring responsibilities, accuracy l of performance data, and use of component monitoring criteria.
! i Service water system vacuum breaker test failures continued in spite of previously , established measures to ensure that valve operability was maintained. Until a design ! change could be developed, PSE&G personnel appropriately devised and implemented an aggressive maintenance and testing plan which improved valve reliabilit The inspectors judged that Station Operations Review Committee reviews were effective and that the committee's charter to perform safety-focused evaluations was fulfille V i l l l ! l
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! Plant Succort i 4-The inspectors observed generally good performance by site security personnel throughout the report period. Additionally, the newly installed " hand geometry" access control system j for the site worked well.
l The con:rol room fire detection and suppression system was installed, maintained, and . j- operated in accordance with applicable design documentation. Operatioa department
- _ personnel were properly trained and qualified to employ the systems available to fight l control room fires.
I PSE&G aporopriately responded to a self identified issue involving inadequate and untimely
- testing of Appendix R emergency lighting. Several additioreal discrepancies were identified
by PSE&G during the follow up investigation which were properly tracked and i dispositioned in accordance with corrective action program, f
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TABLE OF CONTENTS EX EC UTIVE SU M M ARY . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . v i TABLE O F CO NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vil I. Operations..................................................... 1
- 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 01.1 General Observations .......,....................... 1
- 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
- ' 08 Miscellaneous Operations issue ............................. 2 08.1 (Closed) URI Item 50-354/96 80 02: standby liquid control pumps potentially inoperable following inservice testing . . . . . . . 2
C. ' 2 (Closed) LER 50 354/97-014: technical specification prohibited t Condition ........................................ 2 08.3 (Closed) VIO 50 354/96-04-07: failure to staff the offsite and . onsite safety review groups in accordance with technical - specification requirements ............................ 2 [ 08.4 (Closed) VIO 50 354/96-07 02: failure of the offsite safety l review group to review safety evaluations as required by l technical specifications .............................. 2 08.5 (Closed) VIO 50-354/116-09-01: failure of the offsite safety , review group to maintain adequate staffing levels and failure to review proposed changes to technical specifications ......... 3 II. Maintenance................................................... 3 j M1 Conduct of Maintenance .................................. 3 , M 1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
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M1.2 "C" Emergency Diesel Generator Surveillance Testing . . . . . . . . . 4 M7 Quality Assurance in Maintenance Activities .................... 6 M8 Miscellaneous Maintenance issues .....................,,.... 6 i M 8.1 (Closed) URI 50-354/94 13-01: reactor scram during routine i surveillance testing ................................. 6 M8.2 (Open) URI 50 354/97 01-03: Filtration, Recirculation, and ' Ventilation System (FRVS) surveillance testing methodology . . . . 6 M8.3 (Closed) LER 50-354/95-041: engineered safety feature start , of the "D" station service water pump and "B" control room emergency filtration f an . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M8.4 (Closed) LER 50-354/97-009: unplanned high pressure coolant , injection (HPCI) system inoperability ..................... 7 i M8.5 (Closed) LER 50-354/97 15: radioactive effluent samples not analyzed within required surveillance test interval . . . . . . . . . . . . 8 i M8.6 (Closed) LER 50-354/97 016: Filtration, Recirculation, and Ventilation System technical specification surveillance requirement compliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 i M8.7 (Closed) VIO 50-354/94-09-05: loss of spent fuel pool ! inventory ........................................ 8 ' M8.8 (Closed) VIO 50-354/96-10-01: two examples of a failure to . vii
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y i maintain procedures for surveillance and test of safety related equipment ....................................... 8 M8.9 (Closed) VIO 50 354/96-80-03: two examples of safety related maintenance procedures that were either not used or did not provide adequate information for component assembly . . . . . . . . 9 Ill. Engineering ...... ............................................ 9 E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . 9 E2.1 Station Service Water " Maintenance Rule" Implementation . . . . . 9 E2.2 Station Service Water System Vacuum Breakers ........... 10 E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 11 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 E8.1 (Closed) URI 50 354/93 30-01: operating cycle transient monitoring ...................................... 12 E8.2 (Closed) URI 50 354/94 24-01: pressure locking and thermal binding of gate valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 E8.3 (Closed) URI 50 354/96-06-02: apparent loss of reactor building-to-suppression chamber vacuum breaker configuration control......................................... 12 E8.4 (Closed) URI 50 354/96 80-04: safety-related battery electrolyte temperatures . , , . . . . , . . . . . . . . . . . . . . . . . . . 13 E8,5 (Closed) URI 50 354/97 01-01: impact of safety auxiliaries cooling system (SACS) operation at low temperatures ....... 13 E8.6 (Closed) VIO 50 354/96 14-01014: SACS operated at temperatures lower than described by UFSA9 . . . . . . . . . . . . . 14 I V . Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . 14 S1.1 G e neral Com ments - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 F2 Fire Protection Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . 15 F2.1 Control Room Fire Protection ......................... 15 F2.2 Em erge ncy Lig hting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 V. M anag em ent Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 X1 Exit Me eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 X2 Pre-Decisional Enforcement Conference Summary ............... 17 viii
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. 4-Br.gort Details 1. Operations 01 Conduct of Operations QL1 Reneral Observations (71707) The inspectors conducted frequent reviews of ongoing plant operations throughout the report period. Several plant evolutions were observed, including:
* turbine auxiliaries cooling system transfer to alternate supply * reactor wcter cleanup system placed in service following outage * reactor protection system bus transfer o residual heat removal pump inservice testing Overall, the inspectors observed professional and safety-conscious plant operation Good communications, and command and control of routine plant evolutions was evident. Senior operators were generally aware of ongoing plant maintenance and other station activities which affected control room indications. Control room -
operators strictly adhered to procedure guidance. Senior operations department supervision were often in the field monitoring emergent corrective maintenance and troubleshooting activities. Operations-led pre-job briefings were conducted when appropriate and were of high quality: communications, contingencies, and individual responsibilities were discusse Unexpected event response was also evaluated. The inspectors reviewed operator actions during several events, including: -
* reactor manual control system lock up * reactor water cleanup system isolation * TS 3.0.3 entry for both trains of control room effluent filtration inoperable e automatic isolation of turbine auxiliaries cooling system The inspectors verified that following each of the events, appropriate TS action statements were entered and tracked to completion, Accurate and timely reports of applicable non-emer :9ncy events were made to the NRC in accordance with 10 CFR 50.72. The correctiN action program was employed when conditions adverse to quality were identified. The inspectors noted that the number of errors committed by operators was lower than in previous report periods, and showed a steady decline since March 1997. The inspectors attributed this decline to increased emphasis on human error reduction by department management, use of the control room " peer check" process, and enhancement of operating procedures to reduce ambiguit The inspectors concluded that, in spite of frequent challenges from both unexpected and planned events, the operations department exhibited improved performance during the period. Communications, command and control, and compliance with TS requirements were all goo _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ - - _ _ -
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07 Quality Assurance in Operations The inspectors observed frequent quality assurance (QA) department personnel coverage both in the control room and in the field. QA supervisors actively participated in management and station operations review committee meetings and provided objective, independent feedback and insight into current Hope Creek operations. Based on a review of the monthly QA report and discussion with various staff, the inspectors judged that oversight of operations department activities was appropriat Miscellaneous Operations issue 08.1 (Closed) URI ltem 50-354/96-80 Oh standby liquid control pumps potentially inoperable following inservice testing. In February 1996, the NRC Readiness Assessment Team inspection (RATI) noted several failures of standby liquid control (SLC) pumps to develop the technical specification (TS) required flow during inservice testing (IST), the cause(s) of which had not been well understood. The RATI team identifiod the possibility that the SLC pumps had reduced performance due to air binding introduced during performance of the IST. The RATI team left this issue unresolved pending a review for historical inoperability and NRC reportability requirement The inspectors reviewed PSE&G's root cause analysis and follow-up assessment of the repeat IST failures conducted after the RATIidentified the issue. Based on this effort, the inspectors concurred with PSE&G's conclusion that the SLC pumps remaind operable following previous IST's despite the potential introduction of small amounts of air in the suction piping. PSE&G subsequently revised the SLC_ . pump IST procedure to preclude the introduction of any air. The inspectors verified that since the revision, no test failures have been experience .2 (Closed) LER 50-354/97-014: technical specification prohibited condition: failure to , complete offsite power distribution line up within required time frame. This event was discussed in NRC Inspection Report 50-354/97-04 and was dispositioned as a Non Cited Violation. No new issues were revealed by the LE .3 (Closed) VIO 50-354/96-04-07: feiture to staff the offsite and onsite safety review groups in accordance with technical specification requirements. This issue is identical to that described and closed in NRC Inspection Report 50-272 and 50-311/96-15 for PSE&G's Salem station. A similar issue is also described in section 08.5 belo .4 (Closed) VIO 50-354/96-07-02: failure of the offsite safety review group to review safety evaluations as required by technical specifications. The inspectors verified the corrective actions described in PSE&G's response letter, dated November 18, 1996, were reasonable and complete. These actions included a review of the missed safety evaluations, and the conduct of an audit of all evaluations completed in the past three years to ensure no others failed to receive appropriate reviews. A similar issue is described in section 08.5 belo .
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08.5 (Closed) VIO 50 354/96-09 01: failure of the offsite safety review group to maintain adequate staffing levels and failure to review proposed changes to - technical specifications. The inspectors verified the corrective actions described in PSE&G's retoonse letter, dated December 26,1996, to be reasonable and complete. Though the issues identified in this violation were similar to previously cited violations, the causal factors were deemed to be different and therefore prior corrective actions would not necessarily have precluded these subsequent issue Actions in response to this violation included a broad review of offsite safety review activities as part of a self assessment. Additionally, PSE&G recently implemented a new NRC-approved quality assurance program (technical specification amendment 97) which disbanded the offsite safety review group and instituted and new independent review body, the Nuclear Review Board. The inspectors have witnessed the activities of this new board and judged them to be effective. No similar problems have been identified since this violation was issue II. Maintenance M1 Conduct of Maintenance M 1.1 Maintenance Observations a, insoection Scooe (62707) The inspectors observed work activities on severalimportant to safety systems and components, including:
* reactor manual control system
* reactor protection system motor generator and electrical protection assemblies
* safety auxiliaries cooling system isolation valve 1EGHV 2522B
* south plant vent radiation monitoring system
* turbine auxiliaries cooling system low pressure alarm circuitry Observations and Findinos The inspectors focused on maintenance department troubleshooting activities during the report period, and in general observed an appropriate level of department supervision, engineering support, and pre-job planning and briefing for each of the issues. Operating experience was utilized when available, and was particularly useful during the evaluation of a reactor manual control system (RMCS) failure on August 18. A similar RMCS failure occurred in 1996, and an extensive root cause assessment was completed following that event which aided in resolution of the later issue. In two cases, the south plant vent radiation monitor and the reactor protection system electrical protection assemblies, pre-approved hardware design changes, developed as the result of pravious poor system performance or industry information, were available to resolve the equipment failure .
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_ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ v l 4 Maintenance technicians frequently employed troubleshooting procedure HC lC-GP.ZZ-008(O) when pursuing resolution of emergent corrective maintenance issue This procedure permits the use of problem specific troubleshooting plans to analyze system failure mechanisms. The inspectors observed that the quality of these plans varied; initial plans were typically brief and informal; detailed formal plans were not generally developed unless difficulty or delays were experienced in isolating the f ailed component (s). Once formal, detailed plans were established, technicians often identified failed system components readily, For example, repairs of the failed RMCS following initial troubleshooting were short lived; within three hours of the work the system failed again. However, after a detailed work plan was subsequently uaveloped, which was based on input from several disciplines, technicians discovered and replaced a malfunctioning " activity control card" in the unit which permanently corrected the f aul Procedure compliance during the observed work activities was generally good, with one notable exception. Again, while witnessing RMCS work in progress, the inspectors identified a power supply preventative maintenance procedure which had been completed but, the precautions and limitations section of the procedure had not been initialed as being periormed. The inspectors identified this issue in spite of the fact that both maintenance department supervision and management personnel were present at the job site. A failed power supply in the RMCS was later attributed to the technicians who performed the power supply checks. The inspectors judged this incident as an additional example of the violation cited in NRC Inspection Report 50-354/97-03, which documented two examples of failures by maintenance technicians to follow written procedures. Corrective actions for the noted previous violation have not been fully implemented, Conclusions Maintenance technicians appropriately implemented troubleshooting activities following unexpected f ailures of important to safety equipment, however the quality of tha troubleshooting plans varied. Supervision at job sites was evident, however in one case both a supervisor and manager failed to identify missed steps in a preventive maintenance procedure. Good use of both internal and external operating experience, was observe M12 "C" Emeraency Dierel Generator Surveillance Testina Insoection Scone (61726) The inspectors conducted a thorough review of recent tests of the "C" emergency s diesel generator (EDG). Detailed procedure and work order reviews, system walkdowns, operating logs, test observations, and discussions with responsible system engineers were performed in completing this assessment, i l l
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5 $ " Observations and Findinos During the week of August 11,1997, three separate work orders for "C" EDG
- surveillance testing were implemented
- the routine monthly run for TS 4.8.1.1.2.a,
- -a once-per refueling cycle 24 hour run in accordance with TS 4.8.1.1.2.k and an j overspeed trip test of the EDG governor. The inspectors witnessed good i coordination, planning, and execution of the tests. Deficiencies, when identified,
- were appropriately entered into the corrective action program Following an initial l failure of the overspeed trip test (tripped 2 rpm low), the EDG vendor was
, contacted to provide assistance on trip setpoint adjustment. The vendor articulated i an enhanced method for test conduct which was employed successfully. The j inspectors noted that vendor. manual guidance for completing this test was
- insufficient in that it did not provide the detail necessary to ensure satisf actory -
' completion of the test. Maintenance staff appropriately revised the governing test procedure to preclude recurrence of the test failur '
f The inspectors observed that the documentation associated with these three tests l adequately demonstrated compliance with regulatory requirements, though some
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minor discrepancies were noted. For example, both of the first two tests listed above were completed in parallel; i.e. operators took credit for performance of the
- first test while running the second. Though both procedures were completed during
- the single EDG start and run, the inspectors noted that the implementing procedures
- - were not identical, which introduced the potential _for steps to be performed out of proper sequence or leaving an engine component in the wrong configuration following the test. Operations staff initiated a procedure change to resolve the
- noted discrepancie ,
Lastly,-the inspectors identified a step in a test procedure which appeared to - i -"precondition" the EDG for successful operatiun. Specifically, step 5.1.10 of the , monthly surveillance procedure HC.OP-ST.KJ-0003(Q) requires in part that, just l - before the EDG is started, each fuelinjection pump rack be pushed open against its ! spring (and allowed to return) to check for binding. If binding is evident, the ' ! procedure directs that the linkages be wiped down with fuel to correct the problem.
l The inspectors discussed this issue with both operations and engineering personnel, ! who stated that this exact check is performed every week on every EDG as preventive maintenance measure. Additionally, this activity was instituted as a , corrective action from prior experience with " sticking" injection pump racks. While the inspectors agreed that the checks were prudent as a preventive maintenance-activity, these checks were considered inappropriate for implementation just prior to !,_ 'TS surveillance testing. Operations management initiated procedure changes to eliminate this activity from the monthly tests.
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_ _ - _ _ e 6 Conclusions Monthly emergency diesel generator tests were well coordinated, planned and executed. Weaknesses in procedure quality were noted, both in pre start check activities und test implementation methodology. Identified deficiencies, including a
"proceduralized" preconditioning issue, indicated weaknesses in attention to-detail and usag M7 Quality Assurance in Maintenance Activities The inspectors reviewed the QA department monthly report for July 1997 and discussed the resuits with various department staff ard management. Though much broader in scope and depth, QA assessments were generally consistent with NRC inspector evaluations of maintenance department performance. Several good issues were identified, including: 'e inconsistent action request initiation upon discovery of conditions adverse to ,
quality e inadequate procedure use and compliance e infrequent self assessments in planning and scheduling department act!vities a weaknesses in the troubleshooting process Along with the noted individual deficiencies, QA often highlighted areas needing increased management attention based on evaluation of trend information. The inspectors concluded that the QA department was performing frequent and appropriate evaluations of maintenance activities in accordance with the stat.on QA / progra M8 Miscellaneous Maintenance issues M8,1 (Closed) URI 50-354/94-13-01: reactor scram during routine surviillance testing due to f aulty test equipment. This issue was described in detail in Lieinsee Event 1 Report 50 354/94-11 which was reviewed and closed in NRC Insaection Report 50-354/94-19. In that report, the inspectors concurred with PSE&G's assessment of the events and judged the corrective actions to be appropiicte. No similar issues have been identifie MB,2 LQren) URI 50 354/97-01 03: Filtration, Recirculation, and Ventilation System--- IFMVS) surveliiance testing methodology. The inspectors questioned an apparent iiconsistency between the manner in which PSE&G was testing the FRVS system and the technical specification (TS) surveillance requirement. Specifically, PSE&G interpreted the requirament to continuously operate the system "with the heaters and humidity control units ON for ten hours" to mean that these heaters and systems be OPERABLE, implying that heater cycling to maintain humidity below 55% was adequate to demonstrate compliance. This requirement sterrmed directly from NRC Regulatory Guide 1.52, to which PSE&G is committed in the opdated Final Safety Analysis Report (UFSAR), which mirrors the wurds in technical specifications for safety-related heater testing. This issue was left unresolved
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pending further NRC technical review. However, PSE&G concurrently applied for an amendment to the technical specifications to change the surveillance wording from ON to OPERABL Subsequent NRC review of this issue concluded that the requirement for FRVS heaters to be ON should be interpreted literally, and not simply capable of automatic control as PSE&G had judged. As s result, the proposed (TS) amendment was denied on the basis of inadequate justification, in response, PSE&G formally-revised its commitment to Regulatory Guide 1.52 in the UFSAR by providing a detailed and supportable technical basis for only verifying that the heaters are OPERABLE during suweillance tests. Additionally, PSE&G again submitted an amendment request for the noted technical specification with the recently complet*d technical justificatio Finally, PSE&G voluntarily submitted a licensee event report describing the circumstances of this issue (refer to section M8.6 below) The inspectors judged that PSE&G's response to the FRVS testing issues described above were reasonable and timely. Additionally, PSE&G's failure to adhere to the TS testing requirements identified initially had no safety consequences. However, this issue will remain open pending NRC review of the TS amendment submitta MB.3 (Closed) LER 50-354/95-041: engineered safety feature start of the "D" station service water pump and "B" control room cmergency filtration fan during performance of surveillance test. Both of the subject subsystems automatically and unexpectedly restarted after initial test operation was terminated duc Lo unrelated failures. PSE&G attributed the cause of the automatic starts to a failure by control room personnel to evaluate thoroughly the continuation of the surveillance activity, a Loss of Power / Loss of Coolant Accident sequential loading test of the 'D" Emergency Diesel Generator, following the unexpected equipment malfunction Additionally, weaknesses in the procedure guidance for the test were identifie The inspectors reviewed PSE&G's corrective actions and judged them to be ; appropriate to address the stated causal factors. These actions included operator training, procedure revisions, and equipment reoairs. .The failure mechanisms which resulted in the individual subsystem f ailures during the test were well understood and promptly corrected. All stated correctise actions were independently verified to be complet M8.4 (Closed) LER 50 354/97-009: unplanned high pressure coolant injection (HPCI) system inoperability due to minimum flow bypass failure caused by personnel erro This event was described in detailin NRC Inspection Report 50 354/97-03, and resulted in a Notice of Violation for failure to correct previously identified failures by maintenance personnel to adhere to established procedures for work on safety-related equipment. The inspectors verified that the LER accurately described the event and the corrective actions taken both to restore the HPCI system to an operable status, and to minimize the potential for future procedure noncompliance Additionally, PSE&G committed in the LER to conduct a detailed engineering evaluation to establish the potential safety consequences of the event. The inspectors independently verified the accuracy and conclusions of the engineering
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calculations performed in support of this evaluation which determined that, even with the pump minimum flow bypass valve failed open, HPCI was capable of performing its intended safety function under postulated accident condition M8.5 (Closed) LER 50 354/97 15: radioactive effluent samples not analyzed within required surveillance test interval. This LER describes a self identified issue involving routine failures by Hope Creek chemistry personnel to analyze radioactive liquid waste samples within the technical specification (TS required monthly frequency. " Extent of condition" reviews determined that similar f ailures to analyre . . radioactive effluents had occurred following gaseous samples. PSE&G attributed the cause of these failures to an incorrect interpretation of TS surveillance requirements. Correctivo actions included training of affected personnel and revisions to applicable procedures. The inspectors independently verified the accuracy of the information documented in the LER, and ludged the stated causal factors and corrective actions to be reasonable and appropriate. This licensee-identified and corrected violation is being treated as a Non Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev. (NCV 50-354/97-05 01) MB,6 (Closed) LER 50-354/97-016: Filtration, Recirculation, and Venti'ption System technical specification surveillance requirement compliance. The issues described in this voluntarily submitted event report are assessed in section M8.2 above and in NRC Inspection Report 50 354/97 01. No new issues were revealed by this LE M8,7 (Closed) VIO 50-354/94-09 05: loss of spent fuel pool inventory dun to inadequate ' control of maintenance. The events which led to this cited violation included maintenance technicians-working on spent fuel pool gate seats without an approved procedure, and the manipulation of gate seal air supply valves without a tapou Several other issues were also identified which involved inadequate control of maintenance. The inspectors reviewed PSE&G's corrective actions for this violation as documented in a July 18,1994 letter to the NRC. These actions, which included the creation of a new maintenance procedure for work on fuel pool gate seals, the addition of the gate seal air supply valves in the safety tagging computer system,
- and reinforcement of expectations for procedure compliance, were judged to be reasonable and were verified to be complete M8,8 (Closed) VIO 50-354/96-10-01: two examples of a f ailure to maintain procedures-for surveillance and test of safety-related equipment. The first example resulted from a change to a containment leak rate test procedure due to a modification made to containment purge supply valve seats; the test procedure implementation frequency was inappropriately increased to two years from the TS required six months following the valve modification. PSE&G attributed the cause of this occurrence to an impropei decision made by the design change project team regarding the need to apply for a TS change as a result of the modification. The inspet, tors noted that PSE&G implemented appropriate steps to address this issue, which included enhancements to the design change review process and the submittal of a license amendment on May 5,1997 to change the TS surveillance frequency to two year .
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i The second example involved a f ailure to perform standby liquid contrcl system surveillances in accordance with TS requirements. As noted in the cover letter to NRC Inspection Report 50 354/9610, this issue was adequately described in LER
50 354/96 21, and the documented corrective actions were deemed appropriate, j As such, no further action was described in PSE&G's violation response letter No
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further examples were noted.
l M8,9 (Closed) VIO 50 354/96 80-03: two examples of safety related maintenance
- procedures that were either not used or did not provide adequate information for..
{ component assembly. PSE&G attributed the cause of these violations to the mechanical maintenance organization's failure to assess ard correct the j implementation of procedure compliance expectations. Several corrective measures
were taken as a rmit of these issues, including: " stand down" meetir,ge "ith j maintenance technicians to reiterate procedure compliance expectations, weekly
- checks of completed work orders and procedures to verify proper completion, and ;
' creation of a " subject matter expert" program in mechanical maintenance to * enhance ownership of procedures to improve overall quality. The inspectors verified j . that the stated actions were completed as described, and judged them to be l
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appropriate to correct the stated cause. However, as noted in NRC Inspection Report 50-354/97-03 and section M8.4 above, these corrective actions were not lasting in that similar problems were observed approximately one year later. Failure - to prevent repeat occurrences resulted in issuance of a cited violation in report 97-03.
l hl. Enaineerina i
- E2 Engineering Support of Facilities and Equipment
; EL1, Station Service Water " Maintenance Rule" Imolementatior) insoeetion Scone _j37551) ;
The inspectors reviewed all available information associated with 10 CFR 50.65 (" maintenance rule") implementation for the Hope Creek station service water (SSW) system, including:
* SSW performance criteria and goals e corrective action program documents a work order history and maintenance backlog e performance indicators and maintenance rule electronic database e system performance monitoring quarterly reports ,
Additionally, the inspectors walked down major portions of the system, interviewed system managers, and reviewed proposed system design char:ges.
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_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ W 10 Observations and Findinoa Based on the focused SSW review, the inspectors noted generally appropriate implementation of PSE&G's maintenance rule programmatic requirements. The inspectors verified that system performance criteria were established, based on realistic assumptions, and approved by an expert panel. Additionally, since these performance criteria had been exceeded due to higher than expected system ' unavailability, reasonable goalc were developed and approved, and the system was appropriately categorized for performance monitoring in accordance with part a(1) of the regulation. Planned corrective actions to improve overall system operating performance were comprehensive, and involved several design changes, including proposed modifications to discharge strainer elements and traveling screen basket Lastly, since the implementation of previous corrective actions earlier in the operating cycle (i.e. Increased strainer backwash flow, screen spray nozzle enhancements, etc.), the system has operated more reliably during periods of river grass intrusio The inspectors identified some weaknesses in SSW maintenance rule monitorin However, confusion existed regarding responsibility for initiating action requests when performance criteria or goals were exceeded. Based on a review of system unavailability data, the inspectors determined that SSW goals were exceeded in April 1997, but no action request had been initiated identifying this f act in order to establish causal factors. Secondly, the data contained withir; the electronic catabase (the " Cumulus" system) was somewhat inaccurate riue to a lack of specific system knowledge Dy the individuals who input the data. Finally, no consideration for component level monitoring was given despite evidence that poor performance of the SSW traveling screens was largest contributor to SSW system unavailability, Conclusions Implementation of PSE&G's maintenance rule program for the station service water system was generally good in that the system performance was appropriately classified, monitored and evaluated. Weaknesses were evident in individual monitoring responsibilities, accuracy of performance data, and use of component monitoring criteri E22 Station Service Water System Vacuum Breakers Insoection Scone (37551)
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The insoectors reviewed PSE&G's follow up efforts to repeat station service water (SSW) system vacuum breaker failures. Work order history, corrective action program documents, system walkdowns, and discussions with engineering and maintenance personnel were performed while conducting this assessmen Additionally, root cause and corrective actions for similar vacuum breaker issues documented in LER 50 354/96-24 were reviewe i l
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11 Observations and Findings Several failures of SSW loop vacuum breakers, normally-energized solenoid-operated valves, have been experienced at the station during the last year. As noted in NRC Inspection Report 50 354/97-04, PSE&G engineering staff determined, during a root cause analysis performed as an LER 96 24 commitment, that the physical design of the valve was inadequate for its application in the SSW system. Until an improved design could be identified and installed, the valves were p' aced on a reduced interval test schedule (monthly vs. quarterly) to reduce detris buildup on the valve internals, which was judged to be the cause of the repeat feilure However, as evidenced by continued valve failures, monthly testing did not yield the desired results. In response, system engineering and maintenance personnel developed and instituted a moro aggressive action plan to ensure that the valves remained operable until they could be replac;d via a design change. The plan was I developed following consultation with the valve manufacturer. The plan involved cycling each valve every two weeks, followed by a disassembly, clean and inspect, with an additional stroke as a post-maintenance retest. In spite of the increased component unavailability time induced by performing the frequent maintenance, the inspectors concluded that these actions improved the overall reliability and provided reasonable assurance that the valves remained capable of performing their design basis function, Conclusions Service water system vacuum breaker test failures continued in spite of previously established measures to ensure that valve operability was maintained. Until a design change could be developed, PSE&G personnel appropriately devised and implemented an aggressive maintenance and testing plan which improved overall valve reliabilit E7 Quality Assurance in Engineering Activities The inspectors attended portions of two SORC meetings to evaluate the committee's effectiveness at critically reviewing presented information for potential safety issues. Issues assessed at the meetings included a license change request submittal for torus volume and level, and a safety evaluation for a newly proposed method of shutdowa cooling. In both cases, the SORC effectively challenged the proposals from a safety perspective, in the latter case, an action request was generated to address a TS interpretation issue which arose during discussions between SORC members. The inspectors observed, however, that not all members actively participated in the discussions. Despite obvious efforts by the cummittee chairmen to elicit input from all members, two line managers with significant plant operatione experience tended to dominate the review and approval proces The inspectors judged that the SORC reviews of the two noted issues were effective and that the committee's charter to perform safety-focused evaluations was fulfille _ _ _ _ _ _ _ _ _ _ _
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E8 Miscellaneous Engineering issues E (Closed) URI 50 354/93 30-01: operating cycle transient monitoring. The inspectors questioned the manner in which the Hope Creek engineering department was tracking reactor plani thermal cycles (cooldown to heatup to cooldown) and reactor scrams in order to ensure that the plant design cycle limits in TS section 5.7.1 would not be exceeded. At the time the issue was identiflad, the inspectors determined that the transient information had been recorded but that there was no formal procedure or program in place to monitor trends in performanc During this report period the inspectors votified that PSE&G had developed a formal procedute, HC.RA-PR.ZZ-0005(Q) " Thermal Cycle Monitoring," to track the noted design information. Additionally, the inspectors interviewed the engineering personnel responsible for this program and verified that the necessary information was being recorded and analyzed appropriatel E12 [ Closed) URI 50-354/94-24-01: pressure locking and thermal binding of gate valve This issue, which was generic to the nuclear industry and the subject of several NRC information notices and generic letters, identified the need for a detailed NRC review of PSE&G's calculations and analytical methodology to ensure that potentially susceptible motor-actuated valves had adequate actuator thrust to overcome valve pressure locking and/or thermal binding. This detailed assessment was completed and documented in NRC Inspection Report 50 354/96-04, fil (Closed) URI 50-354/96-06-02: apparent loss of reactor building to-suppression chamber vacuum breaker configetation control. This issue involved a discovery by PSE&G petsonnel following repeat surveillance test failures that the installed configuration of the subject safety-related vacuum breakers did not conform to the as-built technical drawings. Specifically, roll pins were not installed in the vacuum breaker hinge shafts as designed to ensure that disc-to-seat alignment was maintained following adjustment. The inspectors questioned whether PSE&G's subsequent installation of the roll pins was a modification (requiring an evaluation) and not simply a repair activity. The issue remained unresolved pending an engineering department investigation into the matte During this report period the inspectors reviewed PSE&G's root cause evaluation of the issue, which was conducted by an independent multi disciplined team, and judged it to be comprehensive and objective. Based on a review of the facts presented, the inspectors concurred with PSE&G's conclusion that the apparent cause of the configuration control deficiency was a failure by the component vendor to install the roll pins during manufacture. The inspectors further judged that installation of these roll pins would not normally be altered following initial installation, and therefore would not be verified following successful initial testin Finally, the inspectors cgreed that PSE&G's installation of the roll pins was justifiable as a repair activit _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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,F.8.4 (Closed) URI 50 35_4/96 80-04: safety-related battery electrolyte temperatures, This issued involved a question raised by NRC inspectors that the TS minimum 125/250VDC battery electrolyte temperature limit of 60 degrees F was not consimont with electrolyte temperature information documented in the UFSAR (77 5 deg ees F), and that the T.S. would allow battery operation outside the design basis, in response to this discrepancy, PSE&G performed a detailed review of the design and licensing basis te verify the validity of the information in the UFSA This rinalysis concluded that the design of the batteries was consistent with standards provided in IEEE 450-1975, a standard to which PSE&G is also committed in the UFSAR, and that the batteries were capable of performing their design basis function at the TS permitted minimum electrolyto temperatur As a result, PSE&G engineering personnel proposed a change to the UFSAR which clarifiad the design basis minimum electrolyte temperature limit. The inspectors e reviewed the safety evaluation that accompanied the UFSAR change, and attended ,
the SORC meeting that assessed and ultimately approved the proposal. The inspectors were satisfied that the identified discrepancy was appropriately dispositioned and that UFSAR change review and approval process functioned properly in this cas Eqd (Closed) URI 50-354/97 01-01: impact of safety auxiliaries cooling system (SACS) operation at low temperatures. This issue stemmed from a failed surveillance test of the "C" train of the FRVS caused by excessive moisture collecting in differential pressure instrument lines. The moisture was attributed to condensation resulting from very low SACS temperatures (i.e. less than 40 degrees F) in the system air cooler. Compensatory measures instituted to reduce the potential for condensation required frequent operator action to maintain SACS temperatures above 55 degrees F. This unresolved item was opened pending the results of a PSE&G effort to determine whether the SACS operating procedure was inappropriately based on an engineering evaluation conducted a year earlier (see section E8.6 below) which may have improperly concluded that the SACS could be safely operated below the 65 degrees F minimum limi During this report period, as part of the follow up corrective actions for the above described event, PSE&G design engineering personnel identified another issue associated with SACS low temperature operation. Specifically, engineering staff concluded that safety-related ventilation system chiller units could fail following a loss of instrument air (LOlA) event with SACS temperatures lower than 55 degrees F. The insoectors noted that PSE&G engineers had not recognized the potential impact of LOlA events on cooling water systems until August 1996, during SSW and ultimate heat sink design basis review The inspectors reviewed the February 1996,10 CFR 50.59 safety evaluation that justified the UFSAR change to allow SACS operation below 65 degrees F, and determined it to be inadequate in that it failed to account for a LOlA scenario. This was a violation of 10 CFR 50.59 in that the consequences of a malfunction previously evaluated in the UFSAR was increased. The inspectors judged that PSE&G's 10 CFR 50.59 review process had significantly improved since the
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February 1996 safety evaluation was prepared, in part the result of corrective actions to other NRC violations. As such, the inspectors concluded that under the revised review process the impact of a LOlA event would likely have been ' evaluated. This licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll B.1 of the NRC Enforcement Polie (NCV 50 354/97-05-02) At the conclusion of the report period, compensatory measures to assure SACS > temperatures would remain above 55 degrees F were stillin place, though were not needed until service water (which cools SACS) temperatures dropped below 55 degrees. The inspectors observed that PSE&G design engineering management was aggressively pursuing ultimate resolution of this issue, and had developed a comprehensive action plan to ensure a timely and satisfactory conclusio ER (Closed) VIO 50-35 V9614-01014: SACS operated at temperatures lower than described by UFSAR The SACS was originally designed and licensed for a minimum heat exchanger outlet temperature of 65 degrees F, but subsequent PSE&G engineering evaluations were completed that justified system operation to as low as 32 degrees F. At the time this issue was iden+ified, the system had been frequently operated at temperatures below 65 degrees F but the limit in the UFSAR was never revised to reflect PSE&G's re evaluation. PSE&G attributed the failure to update the UFSAR in a timely manner to previously known deficiencies in the station's corrective action program. The inspectors reviewed PSE&G's May 7, 1996 letter in response to this violation, and concluded that proposed corrective actions were appropriate and complete. These actions included a thorough review of selected safe shutdown and risk significant systems to determine if any discrepancies.were evident between design basis information and system operating procedures. Additionally, the Hope Creek UFSAR was revised to reflect the updated SACS design basis informatio IV. Plant Support S1 Conduct of Security and Safeguards Activities S1.1 General Comments
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The inspectors observed generally good performance by site security personnel throughout the report period. Proper staffing of the central and secondary alarm stations were observed and compensatory measures were effectively employed when needed. One example of good security performance involved the discovery of site-prohibited materials in the cab of a commercial delivery truck during a sally port inspection. The noted materials were properly confiscated and the driver detained for questioning before the truck was permitted on site. Additionally, the inspectors noted that the newly installed " hand geometry" access control system for the site worked well, with no instances of unauthorized site access reported or identified during the perio .. . . .
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F2 Fire Protection Facilities and Equipment E2d Control Room Fire Protection insoection Scooe (71750) The inspectors evaluated the status of the fire protection system for the Hope Creek control room. The UFSAR, Safety Eveluation Report (SER) and applicable procedures were reviewed, a system walkdown was conducted, and discussions with fire protection and control room personnel were held, Observations and Findinas The Hope Creek control room does not have an automatic fire suppression syste If a fire were to develop, it would be detected by installed monitors and pre-staged portable halon extinguishers would be used to fight the fire. A large capacity, manually-initiated halon system is available which was designed to extinguish a fire that could potentially start inside the control room operating consoles in order for this system to be employed, operators must first connect a hose between the halon storage tanks adjacent to the control room and fittings on the control panels. The inspectors verified that all the necessary hardware was locally available and in good condition. Additionally, operators were judged to be cognizant of when and how to employ the system. However, the ir'spectors determined that there was no formal procedure established for system operation, likely because of system simplicit This latter fact was communicated to Hope Creek management who acknowledged the deficiency and generated an action request in accordance with the corrective action program to address this minor procedural issu Additionally, the inspectors verified that self-contained breathing apparatus (SCBA) were locally available for all operatione department shift personnel in the event of a fire in the control room. The inspectors reviewed training records and determined that all current control room and equipment operators were qualified to use the SCB A' The inspectors determined that the UFSAR, the SER, and cpolicable procedures accurately described the control room fire protection system as it is currently installed, Conclusions The control room fire detection and suppression system was installed, maintained, and operated in accordance with applicable design documentation. Operations department personnel were properly trained and qualified to employ the systems available to fight control room fire _ _ _ _
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EL2 Emeroency Liahtina Insoection Scoce (71750) The inspectors reviewed PSE&G's actions following a self identified discovery that I Hope Creek battery operated emergency lighting, installed in accordance with 10 CFR 50 Appendix R, had exceeded its required surveillance testing interval by over - l one yea Observstions and Findinos During preparations for a Quality Assurance (QA) department audit of site fire protection in July 1997, a OA inspector discovered that the 18 month surveillance test HC.FP ST QB-0070(F), ~8 hour battery-powered emergency light functional test," had not been performed since November 1994. This testing is required by the Hope Creek fire protection program described in UFSAR section 9.5.3.2.2. This issue was promptly reported to station management and to the NRC in accordance with NFP 57 License Condition 2.C 7, and the necessary testing was complete However, over 40% of the emergency lights failed the test, which resulted in the - need for extensive compensatory measures to ensure that equipment needed for safe shutdown could be accessed in the event of a station blackou The inspectors reviewed the adequacy of the surveillance testing and the subsequent compensatory measures and judged both to be acceptabl Additionally, the inspectors noted that a multi-disciplined team assemblud to evaluate this issue, identified several other peripheral fire protection program deficiencies. Examples included discrepancies in maintenance rule tracking, * acceptance limits for battery testing, and storage and testing of replacement batteries. All of the noted issues were tracked in PSE&G's corrective action program; each was properly evaluated with appropriate corrective measures either implemented or planne The inspectors concurred with PSE&G's assessment that the potential safety consequences of this issue were low in that an adequate number of operable flesh;ights were pre-staged throughout-the facility, and that most of the emergency lights passed the test. Of the 40% which failed, most operated for several hours before the associated batteries depleted. This licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll. of the NRC Enforcement Poliev. (NCV 50 354/97 05 03) Conclusions PSE&G appropriately responded to a self identified issue involving inadequate testing of 10 CFR 50 Appendix R required emergency lighting. Several additional discrepancies were surfaced during the follow up investigation, all of which were properly tracked and dispositioned in accordance with the station's corrective action progra . _ _ _ _ _ _ _ _ _ _ _
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17 L V. Manaaement Meetinas X1 Exit Meeting Summary A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that comparos plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable pertions of the UFSAR that related to the areas inspected. The inspecters verified that the UFSAR
. wording was consistent with the observed plant practices, procedures and/or parameter On August 28,1997, the inspectors presented the inspection results to members of licensee management at the conclusion of the report period. The licensee acknowledged the findings presenk!.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie X2 Pre Decisional Enforcement Conference Summary On August 12,1997, a pre-decisional enforcement conference was held at the NRC Region I office to discuss potential enforcernent issues identified in NRC Inspection Report 50-354/97-01. The issues related to apparent violations of 10 CFR 50.59 during tha installation of a residual heat remtval system cross-tie modification. Slides used in the licensee's presentation at the conference ara included as Attachment A to this repor .
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._ 18 INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations-IP 62707; Mainten::nce Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92700: Licensee Event Reports ITEMS OPENED, CLOSED, AND DISCUSSED i
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Opened / Closed 50-354/97-01-03 URI FRVS surveillance testing methodology 50 354/97-05-01 NCV radioactive effluent samples not analyzed within required surveillance interval 50 354/97-05-02 NCV SACS operation at low temperatures 50 354/97 05 03 NCV emergency lighting surveillance testing interval exceeded l Closed 50-354/93 30-01 URI operating cycle transient monitoring l 50 354/94 09-05 VIO loss of spent fuel pool inventory l 50-354/94 13-01 URI reactor scram during surveillance testing due to faulty test equipment 50 354/94-24-01 URI pressure locking and thermal binding of gate valves 50-354/96-04-07 VIO failure to staff onsite and offsite safety review groups 50-354/96-06-02 URI apparent loss of reactor building to suppression chamber vacuum breaker configuration control 50 354/96-07-02 VIO failure of offsite safety review group to review safety evaluations 50-354/96-09-01 VIO failure of offsite safety review group to maintain adequate staffing 50 354/96 10-01 VIO failure to maintain surveillance test procedures 50-354/96-14-01014 VIO SACS operated at temperatures below UFSAR limit 50 354/96 90-02 URI failures of SLC pumps to develop their TS required flow during IST 50 354/96 80-03 VIO - failure to follow procedures 50 354/96 80-04 URI safety related battery electrolyte temperatures 50-354/97-01 01 UR! impact of SACS operation at low temperatures 50-354/95-041 LER engineered safety feature start of the "D" station service water pump and "B" control room emergency filtration fan 50-354/974309 LER unplanned HPCI system inoperability 50-354/97 014 LER failure to complete offsite power distribution lineup _ within TS required time frame 50-354/97 015 LER radioactive effluent samples not analyzed within TS required surveillance interval 50 354/S7-016 LER FRVS TS surveillance requirement compliance a - _ _ _ _ - - - _ _ .
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LIST OF ACRONYMS USED EDG Emergency Diesel Generator FRVS Filtration, Recirculation, and Ventilation System FSAR Updated Final Safety Analysis Report HPCI High Pressure Cootent Injection IST Inservice Testing LER Licensee Event 9 aport LOlA Loss of Instrume... Air NCV Non Cited VinW5n NRC- Nuctsar Reg . rt rcy Commission PDR Public Docu: neat Room PSE&G- Public Service Electric and Gas- - OA Quality Assurance
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RATI Readiness Assessment Team inspection RMCS Reactor Manual Control System SACS Safety Auxiliaries Cooling System SCBA- Self Contained Breathing Apparatus - SER Safety Eval'sation Report SLC Standby Liquid Control SORC Station Operations Review Committee SSW Station Service Water TS Technical Specification URI Unresolved item VIO Violation .
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ATIActerI A The Power of Commitment O PSEG HOPE CREEK GENERATING STATION
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NRC ENFORCEMENTCONFERENCE
l AUGUST 12,1997 MARK BEZILLA l
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. 1 . . i j The Power of Commitment - l .
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PSEG I3TRODUCTION l l . l' PotentialViolations: o Unreviewed Safety Question (USQ) Issue l
' o Inadequate Identification of Surveillance Requirements Safety Perspective:
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o Plant was never in a Condition that Jeopardized the Capability of the RHR System to Perform its Specified ! ' Functions '
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RHRIMPROVEME3TMODIFICATION Reactor Pressure Vessei Primary Containment'
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The Power of Commitment PSEG SAFETY CONSEQUENCES i p Probability per Year of: 1.00E+00 , 1.00E-01 ,_ Simultaneous Failure of a ns 1.00E-02 1.00E-03 - - l l 1.00E-04 .00E-Os Delta l Unintentionally 1.00E-06 ~~ : Mis-positioning Two 1.00E-07 Manual Valves Note: Logarithmic Scale -- _ --
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CONFIGURATION 3/95-12/95 Reactor Pressure Vesse1 Primary Containment'
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CONFIGURATION 12/95-3/96 Reactor Pressure V e s s Primary Containment' r 1P 1P 1P A Reactor JL - JL JL H Q Building x_ X - X I- X
' ' .mencuce MOV h J kvJ P 'T V7 }
taied ChrcJ x S x x A C B ' Pumps J J J zO L 1 C l O I E l l L I k l
~~, .. . , m u m -.,n & .,qd9m 3 m ( + a n s s 4; ,' m j! .o n
,"q+n ta)y>ilyau
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CONFIGURATION 3/96-10/96 Rcaetor Pressure V e. s s e 1 Primary Containment' 1P 1r 1F 1P JL Reactor Jk JL JL H H g_ H X, go;iainy g X M
, -
Unconnectaf I.'namected Mov stOV kJ kv2 * kJ VR VR VR u u ed taked G osed Gosal A .s x x x x A C B D Pumps _ J J J O I O l C C
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CONFIGURATIOK10/96-4/97 Reactor Pressure Vesse1 Prima Containment' s 1F 1P 1P N Reactor JL JL JL N H JL-g Building _ X [X1- X Unconnected Uncwinected Mny MOV Monthly Survei!!ance Initiated
~
LM LeJ L,J kJ 73 73 in Accordance With IS4.6.I. ] 73 73 { LocLed W Led Bad on TSSIP Findine Closed Cksd x x x x - A' C B D Pumps J J J zO C
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CONFIGURATION AFTER4/97 Reactor Pressure .V e s s e 1 Primary Containment' 1F 1P 1P 1P Reactor Jk g H Jk JL M H JL Building X_ X ><3- X 7 tu ,nected Unc. - .oed Mg *v Monthiv Surveillance Being j kI 7 'I hI J Perfo:med in Accordancewith TS Lakel oowa J L ed 1 oded ) 4.6.1. k
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i RHRIMPROVEMENTMODIFICATION i
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1P 1P 1P ' Reactor JL g H JL Jk Dx3_ H go;iainy D><}_ X' -D<l- X;
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Uncowcted Unconnected , Mov Monthly Surveillance Being i
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The Power of Commitment UNREVmWED MMTY PSRG quzsrios s . T o Between April 1994 and March 1995, an USQ Condition existed o Flawed 10CFR50.59 Safety Evaluation Completed in February 1994 o Both Valves Always Closed: No Actual Safety Consequences o Enhancements to 10CFR50.59 Process have been made since the time of the Original Evaluations n -
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ne Power dCommhmem INADEQUATEIDENTIFICATION FSM MI,I,ANCE
: : PSEG REQUIREMENTS i p o issue: Did not identify the need to include Cross-Tie Valves in the monthly ECCS Surveillance o The need to Lock One Valve Closed was Identified by PSE&G in December 1995 o Containment Surveillance issue also identified by-PSE&G o Missed Opportunity to Report Missed Surveiliance and to identify USQ in December 1995 v > : .~ ~ ~
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. . . . . . .
_ _ _ _ _ _ _
. . .
The Power of Commitment
-
10CFR50.59 PROGRAM i _ o Modified Program to be Consist with 4/96 NRC Guidance u Enhanced Training u Established NBU Qualification Program u Evaluation of 50.59 and Engineering Performance Issues / Common Cause Analysis of 50.59 Program a Quality Being Measured a Implementing Independent Peer Review a SORC Training
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ntof10 of ws causeassessme we CF R . 9re 505
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anmon edw causeassess re r oes at
/ ~ m r, Co edu afety f
8 pro c ofs wgm
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nm f rango
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SORCta ed rissu ew alette 96 22,19 o g' First 6050n Aprilv 6 als o
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opo cr N a
[ cec ebyinA P 9 59 qua #4c / miaittedpre nce n *dBU$05 mtoN Fo r
y 8 R Cguidalisheda
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nce+mnce nguidartmguida te esbatsdo ug n NRCM c6c
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g N pestabngtque tarw
9 ** b The Power of Commitment i CONCLUSIONS PSEG
-
i l,
<> Inadequate 10CFR50.59 Evaluation Resulting in USQ -
No Actual Safety Consequences Based on Actual Plant Configuration
'a Surveillance Issue had no Safety Consequences and was self identified with Effective Corrective Actions a 10CFR50.59 Program has been Enhanced i .
_ _ _ _ - _ . . }}