ML20138E420

Safety Evaluation Supporting Util 831107 & 840320 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1 Re post-maint Testing & Reactor Trip Sys Reliability

ML20138E420
Person / Time
Site:	Vermont Yankee
Issue date:	12/02/1985
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20138E382	List: ML20138E378 ML20138E420
References
GL-83-28, NUDOCS 8512130444
Download: ML20138E420 (2)

Text

8 UNITED STATES

. 8 NUCLEAR REGULATORY COMMISSION 3 eE WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GENERIC LETTER 83-28, ITEMS 3.1.1, 3.1.2, 3.2.1, 3.2.2 AND 4.5.1 VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET N0. 50-271

1.0 INTRODUCTION

By letters dated November 7, 1983 and March 23, 1984, the Vemont Yankee Nuclear Power Corporation (the licensee) submitted its responses for the Vemont Yankee Nuclear Power Station to requirements outlined in our Generic Letter 83-28. This review covers Items 3.1.1, 3.1.2, 3.2.1, 3.2.2

• 4.5.1 only.

2.0 BACKGROUND

On February 25, 1983, both of the scram circuit breakers at Unit I 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 after 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 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 (E00), 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 gene:ic implications of the Salem unit incidents are reported in NUREG-1000,

" Generic Implications of the 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 are categorized into four area; (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.

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Items 3.1.1 and 3.1.2 (Post-Maintenance Testing of Reactor Trip System '

(RTS) Components) require that the licensee submit a statement indicatinq that it has reviewed plant test and maintenance procedures and Technical -

Specifications to ensure that post-maintenance operability testing of safety-related components in the RTS is conducted. Also, the licensee should verify that vendor recommended test guidance has been reviewed, evaluated, and where appropriate, included in the test and maintenance procedures or the Technical Specifications.

Items 3.2.1 and 3.2.2 (Post-Maintenance Testing of All Other Safety-Related Components) require that the licensee submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to assure that post-maintenance operability testing of all safety-related equipment is conducted. Also, the licensee should verify that vendor-recommended test guidance has been reviewed, evaluated, and where appropriate, included in the Technical Specifications. -

Item 4.5.1 (RTS Reliability) requires that the licensee provide a commitment to independent, on-line functional testing of the diverse trip features.

3.0 EVALUATION ,

The staff has evaluated the licensee's submittals responding to the '

requirements of Items 3.1.1, 3.1.2, 3.2.1, 3.2.2 and 4.5.1. The staff concludes that the licensee has responded to NRC quidance and provided a s statement that it has reviewed the plant test and maintenance procedures and 3 Technical Specifications and is in compliance with the requirements of Items 3.1.1, 3.1.2, 3.2.1 and 3.2.2. The staff also concludes that the licensee's Technical Specification requirements to perform on-line functional testing of the reactor protection system, including independent testing of,the scram pilot valves, is adequate to meet the requirements of item 4.5.1.

The licensee's justifications for not making the necessary modifications to permit on-line functinnal testing of the backup scram valves will be '

reviewed separately by the staff in a later action as part of Item 4.5.2. L

4.0 CONCLUSION

Based upon the foregoing discussions, the staff concludes that the progrems 's outlined in the licensee's submittals adequately address the requirements of Items 3.1.1, 3.1.2, 3.2.1, 3.2.2 and 4.5.1 of Generic Letter 83-28, atld its responses for the subject items are acceptable. c)

Principal Contributor: Thomas'B. Silko and Kenneth E. Johnston 9

Dated: December 2, 1985 g

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