ML20203N222
| ML20203N222 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/12/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203N218 | List: |
| References | |
| GL-83-28, NUDOCS 8609230222 | |
| Download: ML20203N222 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF THE NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 83-28, ITEMS 3.1.1, 3.1.2, 3.2.1, 3.2.2 AND 4.5.1 PUBLIC 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 Unit 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 tripped manually by the operator about 30 seconds af ter the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 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 Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem incidents are reported in NUREG-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 Reactor Regulation requested (by Generic Letter 83-28 dated' July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas:
(1) Post-Trip Review, (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip System (RTS)
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 3.1.2, 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 Reliability (System Functional Testing) 0FFICIAL RECORD COPY OL HC SE - 0003.0.0 06/10/86 8609230222 860612 PDR ADOCK 05000354 P PDR
1 Safety Evaluation 2 By letter dated December 17, 1984, Public Service Electric and Gas Company (PSE&G-Licensee) described their planned and completed actions regarding
.the above items for Hope Creek Generating Station (HCGS). Certain of these actions were reviewed during a Region I inspection conducted on December 30, 1985 through January 3,1986 (Inspection Report 50-354/85-66).
2.0 Evaluation 2.1 General Generic Letter 83-28 included various NRC staff positions regarding the specific actions to be taken by operating reactor licensees and operating license applicants. The Generic Letter 83-28 positions and discussions of lisensee 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.
2.2 Actions 3.1.1 and 3.1.2, Post-Maintenance Testing (Reactor Trip System Components), and Actions 3.2.1 and 3.2.2, Post-Maintenance Testing (All Other Safety-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 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 (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 documenting 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 components) to assure that any appropriate tests guidance is included in the test and maintenance procedures or the Technical Specifica-tions, where required.
OFFICIAL RECORD COPY OL HC SE - 0004.0.0 06/10/86
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. Safety Evaluation 3 Discussion In the letter dated December 17, 1984, the licensee stated that the post-maintenance testing of safety-related components including reac-tor trip system components at HCGS is conducted in accordance with the Station Administrative Procedure SA-AP.ZZ-009(Q). 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 following design basis events to insure the integrity 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 scheduling, prioritization, maintenance performance and documentation.
The Station Administrative Procedure SA-AP.ZZ-015(Q) establishes control and responsibilities for the station tagging program. l 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 being 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 a'nd equipment. 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).
A review of selected procedures indicated that the licensee has established adequate procedural controls for the conduct of the 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 serification of compliance to acceptance criteria, accuracy and completeness of surveillance test results.
OFFICIAL RECORD COPY OL HC SE - 0005.0.0 06/10/86
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Safety Evaluation 4 The station procedure SA-AP.ZZ-010(Q) contains a mandatory preventive maintenance requirement for all safety-related systems and components. The station preventive maintenance program, conducted to optimize equipment reliability and enhance plant availability, is implemented through the use of the Inspection Order (IO) system. The procedure defines the control and responsibilities for review and evaluation of adequacy 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 judgment.
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 equipment, its maintenance, periodic testing, calibration and design modifications. The licensee's internal vendor interface program is updated by INP0 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 Project Manual Section 1.6, " Response Coor-dination Procedure", including review of the vendors and engineering recocnmendations.
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-040(Q), Master Equipment List; SA-AP.ZZ-047(Q), Operation Experience Evaluation; SA-AP.ZZ-048(Q),
Station Performance and Reliability Monitoring; SA-AP.ZZ-050(Q),
Station Retest Program (Draf t); the Site Engineering Instructions; SEI-2.1, Component Functional Classification; SEI-2.7, 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-maintenance testing, responsibilities, review and approval authority and methods for performing the test f- all safety-related equipment including the reactor trip system components. The licensee's actions to implement the surveillance test procedures are adequate.
OFFICIAL RECORD COPY OL HC SE - 0006.0.0 06/10/86
Safety Evaluation 5 Based 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 83-28 and, therefore, acceptable.
2.3 . Action 4.5.1, Reactor Trip System Reliability (System Functional Testing)
Position 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 undervoltage and shunt trip features on Westinghouse, B&W 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 requirements are to be incorporated in Technical Specifications.
Discussion In the letter dated December 17, 1984, the licensee stated that HCGS performs on-line functional testing of the scram logic channel con-sistent with the proposed Technical Specification requirements. The issued Technical Specification paragraph 4.3.1.1 requires that opera-bility of each reactor protection system instrumentation channel shall be demonstrated by performing 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 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 testing. However, each backup scram valve will be independently tested during each refueling outage. Procedure OP-FT.SB-001(Q), Backup Scram Valve Test was developed to provide the necessary administrative control for these tests. The adequacy of this approach will be reviewed separately under Item 4.5.2 of GL 83-28.
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 83-28 and, therefore, acceptable.
OFFICIAL RECORD COPY OL HC SE - 0007.0.0 06/10/86
, Safety Evaluation 6 3.0 Conclusion Based upon the foregoing 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.
Dated:
Principal Contributor: Madan Dev, Division of Reactor Safety, Region I l
0FFICIAL RECORD COPY OL HC SE - 0008.0.0 06/10/86
, SALP INPUT Facility: Hope Creek Generating Station Docket No.: 50-354 TAC Nos.: None-
. Requested Date: October 30, 1985; Memorandum Novak to Kane Functional Area: Adequacy of Licensee Submittal COMPARISON OF PERFORMANCE TO CRITERIA ATTRIBUTES Criterion'1: Management Involvement and Control in Assuring Quality The licensee has established a Response Coordination Team which is responsible
-for review, approval and implementation of all vendor supplied information, re-gulatory bulletins and instructions, industry standards, engineering recommenda-tions and operational experience as applicable to post-maintenance testing of reactor trip breaker components and other safety-related system and components at HCGS.
Based on the above the licensee involvement and control in assuring quality is rated as Category 1.
Criterion 2: Approach to Resolution of Technical Issues from a Safety Stand-point The licensee has developed and implemented a Station Maintenance Program, Vendor Interface Program, Station Master Equipment List Program, Operational Experience Evaluation Program, Station Performance and Reliability Monitoring-Program, Station Retest Program, and a Component Functional Classification Instruction to support the station approach to resolve technical issues from a safety standpoint relative to Salem ATWS event followup.
Based on the above, the licensee performance is rated as Category 1.
Criterion 3: Response to NRC Initiatives In response to the NRC initiatives, the licensee initiated a Commitment Management Notice, a Reliability and Assessment Management Procedure, and an Engineering and Construction Department Project Manual. The licensee's implementation of the Technical Specification related tests and surveillances for the plant safety-related systems and equipment is underway.
Based on the above, the licensee's response to NRC initiatives to Salem ATWS Event is considered Category 1.
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.. LDRS SALP Input
Conclusion:
- The. licensee's submittal for response to Generic Letter 83-28, Salem ATWS event was found to be adequately stated and enabled a clear understanding of the-technical issues. The licensee's efforts to resolve staff questions concerning the issues were satisfactory.
Rating: Category 1 i
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