ML17254A561
| ML17254A561 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/01/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17254A560 | List: |
| References | |
| GL-83-28, NUDOCS 8510030016 | |
| Download: ML17254A561 (2) | |
Text
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 A ~
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIOH GENERIC LETTER 83-28, ITEMS 3. 1. 3 AND 3.2. 3 (POST-MAINTENANCE TESTING OF REACTOR TRIP SYSTEM)
R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
On July 8, 1983, Generic Letter No. 83-28 was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses, and hol.ders of construction permits.
This letter included required actions based on generic implications of the Salem ATWS events.
These requirements have been published in Volume 2 of NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant".
This evaluation documents the EGSG Idaho, Inc., review of the submittal from Ginna for conformance to items 3. 1.3 and 3.2.3 of Generic Letter 83-28.
EGSG Inc., generated a Technical Evaluation Report (TER) on these issues.
The TER prepared by EGRG Idaho, Inc. contained evaluations for several facilities.
This evaluation has been generated based on the results contained in that TER.
The R.
E. Ginna Nuclear Power Plant is an operating Westinghouse PWR reactor that utilizes a dry containment
- system, two class lE Power Svstem Trains, and a relay type plant protection system logic.
2.0 REVIEW GUIDELINES Item 3. 1.3 (Post-Maintenance Testing of Reactor Trip System Components) requires licensees and applicants to identify, if applicable, any post-maintenance test requirements for the Peactor Trip System (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 safety-related components.
The submittal from Ginna was reviewed to determine compliance with items
- 3. 1.3 and 3.2.3 of the Generic Letter.
The submittal was reviewed to determine that these two items were specifically addressed.
- Second, the submittal was checked to determine if any post-maintenance test items specified in the technical specifications were identified 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.
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The licensee response indicated that there had been no items ideatified in its technical specifications relating to post-maintenance testing that could be demonstrated to degrade rather than enhance safety.
3.0 EVALUATION AND CONCLUSION This safety evaluation was prepared by the NRC staff using input from our contractor, EGSG Idaho, Inc.
Rochester Gas and Electric Corporation, the licensee for the Ginna plant, provided a response to items 3.I.3 and 3.2.3 of Generic Letter 83-28 in its submittal dated November 4, 1983.
The licensee's response to the two items is combined in a single statement that no post-maintenance test'requirements in the existing technical specifications have been identified that degrade rather than enhance safety.
Based on our
- review, we conclude that the licensee's responses to items 3. 1.3 and 3.2.3 for the Ginna Plant are acceptable.
4.0 ACKNOMLEDGEMENT This safety evaluation was prepared by Janet Kelly.
Dated:
October I, 1985