ML20132E628

Safety Evaluation Finding Util 831104 Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing Acceptable

ML20132E628
Person / Time
Site:	Peach Bottom
Issue date:	07/22/1985
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20132E591	List: ML20132E588 ML20132E628
References
GL-83-28, NUDOCS 8508020049
Download: ML20132E628 (2)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO SALEM ATWS EVENT, ITEMS 3.1.3 AND 3.2.3 PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-277 AND 50-278

1.0 INTRODUCTION

By letter dated November 4,1983, the Philadelphia Electric Company (PEco, the licensee) submitted a response to our Generic Letter 83-28 for the Peach Bottom Atomic Power Station, Units 2 and 3. This review covered Items 3.1.3 and 3.2.3.

2.0 RACKGROUND 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 l about 30 seconds after the initiation of the automatic trip signal. The l failure of the circuit breakers has been determined to be related to the j sticking of the under voltage trip attachment. Prior to this incident, on .

February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was cenerated based on 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 I (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 unit incidents are reported in NUREG-1000. " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC) 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-MaintenanceTesting,and(4)ReactorTripSystemReliability improvements.

Item 3.1.3 Components)(Post-MaintenanceTestingofReactorTripSystem(RTS) requires licensees and applicants to identify, if applicable, any post maintenance test requirements for the RTS in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety. Item 3.2.3 extends this same requirement to include all other

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l safety-related components. Any proposed Technical Specification changes resulting from this action shall receive a pre-implementation review by NRC.

! 3.0 EVALVATION Our review of the licensee's submittals was performed with the assistance of EGAG, Idaho, Inc. The submittal from PEco was reviewed to determine compliance with items 3.1.3 and 3.2.3 of the generic letter. First, the submittal was reviewed to determine if these two items were specifically addressed. Second, the submittal was checked to determine if there were any post-maintenance test requirements specified by the Technical Specifications that were suspected to degrade rather than enhance safety.

Last, the submittal was reviewed for evidence of special conditions or other significant information relating to the two items of concern.

The review of Generic Letter 83-73, Item 4.5.3 may result in proposed changes to the Technical Specifications reouirements for surveillance testing frequency and out-of-service intervals for testing. The primary concern of Item 4.5.3 is the surveillance testing intervals. Items 3.1.3 and 3.2.3 are specifically directed at post-maintenance test requirements.

These concerns are essentially independent. However, the evaluation of these concerns are coordinated so that any correlation between these concerns will be adequately considered. Since no specific proposal to change the Technical Specifications has been submitted, there is no l identifiable need at this time for correlating the reviews of item 4.5.3 l with this review.  !

We have reviewed the November 4, 1983 PEco response to items 3.1.3 and 3.2.3 of Generic Letter 83-28. Within the response, the licensee's evaluation for Items 3.1.3 and 3.2.3 is that, following a review of the Peach Bottom Technical Specifications, no existing testing requirements in the Technical Specifications which degrade safety in the Reactor Protection System or other safety-related components were identified.

4.0 CONCLUSION

The licensee stated that it has reviewed its Technical Specification requirements to identify any post-maintenance testing which could be demonstrated to degrade rather thhn enhance s'afety and found none that degraded safety. Based on our review, assisted by our contractor, EG&G, Idaho, Inc., we find that the licensee's submittal, with respect to the Peach Bottom facilities, is acceptable.

Principal Contributor: D. Lasher Dated: July 22, 1985 L