ML20206R069

Safety Evaluation Supporting Util 841217 Response to Generic Ltr 83-28,Items 3.1.1,3.2.1,3.2.2 & 4.5.1 Re Required Actions Based on Generic Implications of Salem ATWS Events

ML20206R069
Person / Time
Site:	Hope Creek
Issue date:	06/25/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20206R065	List: ML20206R060 ML20206R069
References
GL-83-28, NUDOCS 8607070062
Download: ML20206R069 (6)

Text

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7; :j WASHINGTON, D. C. 20555

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SAFETY EVALIIATION RY THE OFFICE OF THE NtJCLEAR REACTnD REGt!LATION RELATED TO GENFRIC LETTER 83-?8, ITEMS 3.1.1, 3.1.2, 3.2.1, 3.?.2 AND 4.5.1 Pl1BLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK GENERATING STATION DOCKET NO. 50-354 1.0 Introduction On February 25, 1983, both of the scram circuit breakers at lJnit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant startup, and the reactor was trioned manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the stickinq of the undervoltace trip attachment. Prior to this incident, on February 22, 1983, at linit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated due to a steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic implications of these occurrences at tinit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the ceneric implications of the Salem incidents are reported in NtJREG-1000,

" Generic Implications of ATWS Events at the Salem Nuclear Power Plant."

As a result of this investigation, the Director, Division of Licensing, Office of Nuclear Deactor Pegulation reouested (by Generic letter 83-28 dated .luly 8, 1983) all licensees of operating reactors, aoolicants for an operatino license, and holders of construction permits to respond to

certain generic concerns. These concerns are categorized into four areas:

! (1) Post-Trip Review, (2) Ecuipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Peactor Trio System (RTS) i Reliability Improvements. Within each of these areas, various specific actions were delineated.

This safety evaluation (SE) addresses the following actions of Generic letter 83-28: ,

-- 3.1.1 and 3.1.7, Post Maintenance Testing (Reactor Trip System Components)

-- 3.2.1 and 3.2.2, Post Maintenance Testing (All Other Safety-Related Components)

-- 4.5.1, Reactor Trip System Deliability (System Functional Testing) 8607070062 860625' PDR ADOCK 05000354 A PDR.

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By letter dated December 17, 1984, Public Service Electric and Gas Company (PSEAG-Licenseel described their planned and completed actions reoarding the above items for Hone Creek Generating Station (HCGS). Certain of these actions were reviewed during a Region I inspection conducted on December 30, 1985 throuah ianuary l 3, 1986 (Inspection Report 50-354/85-66).

2.0 Evaluation 2.1 General Generic Letter 83-28 included various NRC staff positions regardina the specific actions to be taken by operating reactor licensees and operating license applicants. The Generic letter 83-28 positions and discussions of licensee compliance regarding Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2 and 4.5.1 for HCGS are presented in the sections that follow.

?.2 Actions 3.1.1 and 3.1.?, Post-Maintenance Testino (Reactor Trin System Components), and Actions 3.2.1 and 3.2.2, Post-Maintenance Testinn (All Other Safetv-Related Components)

Positions Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system (RTS) is required to be con-ducted and that the testing demonstrates that the eouipment is capable of performing its safety functions betore being returned to service.

Licensees and apolicants shall submit the results of their check of vendor and engineering recommendations (regarding safety-related components in the RTS) to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

Licensees and applicants shall submit a report documentino the extending of test and maintenance procedures and Technical Specif-ications review to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Licensees and applicants shall submit the results of their check of .

vendor and engineering recommendations (all other safety-related i components) to assure that any appropriate tests cuidance is included in the test and maintenance procedures or the Technical Specifica-tions, where required.

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Discussion In the letter dated December 17, 1984, the licensee stated that the post-maintenance testing of safety-related components including reac-tor trip systen components at HCGS is conducted in accordance with the Station Administrative Procedure SA-AP.ZZ-009f0). The licensee has established safety-related functional classification for each HCGS structure, system and component. The safety-related designation applies to that equipment which is required to remain functional during and followino design basis events to insure the intecrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, and prevent or mitigate the consequences of accident that could result in potential offsite radiation exposure. The above procedure designates the responsibility for insuring that the post-maintenance testing of the reactor trip breaker components as well as other safety-related components is conducted to demonstrate that the equipment is capable of performing its intended safety function prior to being returned to service. The procedure delineates equipment maintenance programs, including maintenance activities, planning and schedulino, prioritization, maintenance performance and documentation.

The Station Administrative Procedure SA-AP.ZZ-015f0) establishes control and responsibilities for the station tagging program.

Accordingly, an independent verification by qualified station personnel is required for safety-related station equipment removed from service for maintenance or repair and prior to heino restored to service, including verification of the Technical Specification related action statements, performance and review of retests, and test results, to determine the operability of the systems and equipnent. The NRC Region I inspection (Inspection Report 50-354/85-66) verified that the licensee has established control and responsibilities for the Technical Specification surveillances in accordance with the Station Administrative Procedure SA-AP.ZZ-012(Q).

l A review of selected procedures indicated that the licensee has i established adequate procedural controls for the conduct of the l Technical Specifications related surveillance tests. These procedures contained adequate acceptance criteria, precautions and prerequisites and were written to assure satisfactory performance between surveillances. The procedures also required an independent verification for restoration of components such as valves, breakers, fuses, switches and relays by a responsible individual other than the person performing the activities. In addition, review and approval of surveillance activities by the cognizant individuals provided verification of compliance to acceptance criteria, accuracy and completeness of surveillance test results.

1 The station procedure SA-AP.ZZ-010(0) contains a mandatory preventive niintenance requirement for all safety-related systems and 1 components. The station preventive maintenance program, conducted to optimize equipment reliability and enhance plant availability, is implemented through the use of the Inspection Order (TO) system. The procedure defines the control and responsibilities for review and evaluation of adecuacy of the preventive maintenance program as well as any changes thereto. The program evaluation is based on vendor recommendations, regulatory requirements, operating experience, equipment maintenance history and engineering ,iudonent.

The licensee reviews the vendor documents and technical manuals in accordance with the Site Engineering Instruction, SEI-5.3. This pro-cedure establishes control and responsibilities for review of vendor supplied information to determine its adequacy and applicability to the plant specific eouipment, its maintenance, periodic testing, calibration and design mc?ifications. The licensee's internal vendor interface program is updated by INPO NUTAC Vendor Equipment Technical Information Program (VETIP) and uses Nuclear Plant Reliability Data System (NPRDS) to monitor reliability of selected plant systems and components. In order to properly evaluate and implement recommenda-tions from the NSSS supplier, vendors, industry, and regulatory agencies, HCGS has established a Response Coordination Team (RCT).

This team carries specific authority and responsibilities as described in the Reliability and Assessment Management Policy M3-POP-003 and the HCGS Proiect Manual Section 1.6, " Response Coor-dilation Procedure", including review of the vendors and engineering recommendations.

In addition, in order to accommodate the review, evaluation and conduct of preventive maintenance and testing, the licensee has developed and employed several procedures, such as Station Administrative Procedures SA-AP.ZZ-040f0), Master Eouipment list; SA-AP.ZZ-047I0), Operation Experience Evaluation; SA-AP.ZZ-048(0),

Station Performance and Reliability Monitoring; SA-AP.ZZ-050f0),

Station Petest Program (Draft); the Site Engineering Instructions; SEI-2.1, Component Functional Classification; SEI-?.7, System Analysis Group-Response to Nuclear Industry Documents Coordinated by Response Coordination Team; SEI-5.1, Hope Creek Master Equipment List; and SEI-5.10, Hope Creek Master Equipment List Input / Update. A review of the station surveillance test procedures and instructions indicated that the licensee has adequately addressed criteria for post-mair,tenance testing, responsibilities, review and approval authority and methods for performing the test for all safety-related eouipment including the reactor trip system components. The licensee's actions to implement the surveillance test procedures are adequate.

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Rased on the above, the staff concludes that the licensee's actions are consistent with the NRC staff positions for Actions 3.1.1, 3.1.2, 3.2.1 and 3.2.2 of Generic letter 83-78 and, therefore, acceptable.

?.3 Action 4.5.1, Reactor Trip System Reliability (System Functional Testina)

Pccition On-line functional testing of the reactor trip system, including inde-pendent testing of the diverse trip features, shall be performed on all plants. The diverse trip features to be tested include the break-er undervoltaae and shunt trip features on Westinohouse, RAW and CE plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants; and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

The reauirements are to be incorporated in Technical Specifications.

Discussion In the letter dated December 17, 1984, the licensee stated that NCGS performs on-line functional testinp of the scram lopic channel con-sistent with the proposed Technical Specification reouirements. The issued Technical Specification paragraph 4.3.1.1 requires that opera-bility of each reactor protection system instrumentation channel shall be demonstrated by performino channel check, channel functional test and channel calibration at specified frequencies. The Technical Specification also requires that the logic system functional tests and simulated automatic operation of all channels be performed at least once per 18 months. The licensee has developed and implemented 4 the Hope Creek Generating Station Technical Specification Surveillance Procedure System Cross Reference Matrix Report. This matrix documents surveillance test frequencies,: operational conditions, and the applicable procedures.

The licensee does not intend to test the backup scram valves as part of the on-line functional testina. However, each backup scram valve will be independently tested during each refueling outape. Procedure OP-FT.SB-001(0), Backup Scram Valve Test was developed to provide the necessary administrative control for these tests. The adequacy of this aoproach will be reviewed separately under Item 4.5.2 of GL 83-78.

With the exception noted in the above paragraph, the staff concludes that the licensee's actions in this regard are consistent with the NRC staff position for Action 4.5.1 of Generic letter 83-78 and, I therefore, acceptable. l

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3.0 Conclusion Based upon the forecoing discussions, the staff concludes that the licen-see has complied with Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2, and 4.5.1 of Generic letter 83-28.

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