ML20134K735

Safety Evaluation Accepting 831107 Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing

ML20134K735
Person / Time
Site:	Vermont Yankee
Issue date:	08/19/1985
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20134K731	List: ML20134K727 ML20134K735
References
GL-83-28, NUDOCS 8508300388
Download: ML20134K735 (2)

Text

. , _ . . . _ - -. . -- .

sanw 4 UNITED STATES

[ NUCLEAR REGULATORY COMMISSION i 3 WASHINGTON, D. C. 20555 1

%...../;C SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO SALEM ATWS EVENT, ITEMS 3.1.3 AND 3.2.3 -

-VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 1

1.0 INTRODUCTION

f i

7_ . By letter dated November 7,1983, the Yemont Yankee Nuclear Power Corporation (the licensee) submitted a response to our Generic Letter 83-28 for the Yemont Yankee Nuclear Power Station. This review covers Items' 3.1.3 and 3.2.3.

t0 BACKGROUNP

~

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 after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking o,. of the under voltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an z.utomatic -

trip signal was generated based on steam generator low-low level during plant startup.- In this case, the reactor was tripped manual,1y 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

• .7 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 the 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-Maintenance Testing, and (4) Reactor Trip System Reliability Improvements. .

Item 3.1.3 (Post-Maintenance Testing of Reactor Trip System (RT()

Components) requires licensees and applicants to identify, if applicable, any post-maintenance for the RTS in existing Technical Specificalions which can be demonstrated to degrade rather than enhance safety. Item 3.2.3 extends this same requirement to include all other safety-related 8508300308 850019 PDR ADOCK 05000271

.P. PDR

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

3.0 EVALUATION Our review of the licensee's submittal was performed with the assistance of EG&G, Idaho, Inc. The submittal from the licensee 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 83-28, Item 4.5.3 may result in proposed changes to the Technical Specification requirements 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. 14owever, 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 identifiable need at this time for correlating the reviews of Item 4.5.3

. with this review.

We have reviewed the November 7, 1983 licensee response to Items 3.1.3

- and 3.2.3 of Generic Letter 83-28. Within the response, the licensee's

~

evaluation is that at the present time there are no proposed Technical  :

Specification changes required for Items 3.1.3 and 3.2.3. The licensee has

~

committed to continuously review Technical Specification requirements and - -

. . identify the need for revision in the future. A complete Technical ,

-+ Specification review to clarify wording and provide consistency with all 1 Technical Specification sections has been planned. Any identifiable J deficiencies would be the subject of future proposed Technical l Specification changes. I 1

4.0 CONCLUSION

l The licensee stated that no items have been identified in the Technical Specifications that degrade safety. Based on our review, assisted by our 1 contractor, EG&G, Idaho, Inc., we find that the licensee's submittal is I acceptable. ',

Principal Contributor: D. Lasher  :. I Cf Dated: August 19, 1985