05000219/LER-1982-020, Forwards LER 82-020/01T-0.Detailed Event Analysis Encl

Forwards LER 82-020/01T-0.Detailed Event Analysis Encl

ML20052B090
Person / Time
Site:	Oyster Creek
Issue date:	04/16/1982
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20052B091	List: 05000219/LER-1982-020, Forwards LER 82-020/01T-0.Detailed Event Analysis Encl 05000219/LER-1982-020-01, /01T-0:on 820328,MSIV Leak Rate Testing Disclosed Deterioration of Valve Internals on NS03-A & Packing Leak on Outboard Valve NS04-A.Caused Not stated.NS03-A Disassembled & repaired.NS04-A Repacked
References
NUDOCS 8204300011
Download: ML20052B090 (3)

LER-1982-020, Forwards LER 82-020/01T-0.Detailed Event Analysis Encl

Event date:
Report date:
2191982020R00 - NRC Website

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P.O. Box 388 Forked River, New Jersey 08731 609-693-6000 Writer's Direct Dial Number:

April 16, 1982 CD o>

Y Mr. Ibnald C. Haynes, Administrator 3...

Region I o?

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U.S. Nuclear Regulatory Omnission JR 6

631 Park Avenue

'i King of Prussia, PA 19406

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Dear Mr. Haynes:

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Subject: Oyster Creek Nuclear Generating Station Ch Docket No. 50-219 Licensee Event Report Reportable Occurrence No. 50-219/82-20/0lT

'Ihis letter forwards three copies of a Licensee Event Report to report Reportable Occurrence No. 50-219/82-20/0lT in certpliance with paragraph 6.9.2.a.3 of the Technical Specifications.

Very truly yours, Peter B. Fiedler Vice President & Director Oyster Creek PBF/kdk Enclosures cc: Director (40)

Office of Inspection and Enforcenent U.S. Nuclear Regulatory Catmission Washington, D. C.

20555 Director (3)

Office of Management Information and Program Control U.S. Nuclear Regulatory Comnission Washirgton, D. C.

20555 NRC Resident Inspector (1)

Oyster Creek Nuclear Generating Station Ebrked River, N. J.

08731 8204300011 f s>?

GPU Nuclear is a part of the General Pubhc Utihties System

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OYSTER CREEK NUCLEAR GENERATING STATION Forked River, New Jersery 08731 Licensee Event Report Reportable Occurrence No. 50-219/82-20/01T Report Date April 16, 1982 Occurrence Date April 1,1982 Identification of Occurrence It was identified that an abnormal degradation of the primary containment existed based on the results of leak rate testing performed on the Main Steam 12clation Valves NS03A and NSO4A.

TFis event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9.2.a.3 Conditions Prict to Occurrence The reactor was in cold shutdown condition at the time the occurrence was identified, with the reactor coolant temperature less than 2120F and the reactor vented. The reactor was in various operating modes since the last surveillance was performed satisfactorily.

Description of Occurrence On February 8,1982, while performing local leak rate testing on Main Steam Isolation Valves, the leak rate for valve NS03A was found to be 100 SCFH, which was greater than the acceptable limit of 11.9 SCFH.

On March 28, 1982, the other Main Steam Isolation Valve in the "A" line, NSO4A, was leak tested and resulted in a measured leakage of 22.9 SCFH, due to a packing leak.

In the "A" line piping configuration, a packing leak in the NSO4A and excessive leakage through NS03A provides an abnormal flow path from the reactor vessel to the trunnion room, which is considered part of Secondary Containment.

Apparent Cause of Occurrence NS0 3A:

The cause of the excessive valve leakage is attributed to a valve poppet pad wearing on the lower valve body rib. The body rib, which acts as an alignment guide for the valve poppet, prevented the poppet

- from properly positioning itself into the valve seat during closing.

The cause of the valve deterioration is not fully understood at this time and will be the subject of an engineering evaluation.

Licensee Event Report Page 2 Reportable Occurrence No. 50-219/82-20/01T Apparent Cause of Occurrence (Continued)

NSO4A:

The cause is attributed to' packing leakage.

Analysis of Occurrence As indicated in the Apparent Cause of Occurrence section, the leakage path for NSO4A was through the valve packing.

Because NSO4A is downstream of NS03A, the leakage t te from primary containment was limited to 22.9 SCFH and confined within che trunnion room (part of Secondary Containment). The trunnion room is closed and sealed during power operation to maintain secondary containment.

During power operation there were no indications of abnormal trunnion room temperatures which confirms the leakage to seconda.ry containment from NSO4A was minimal.

Based on the above,' the significance of this occurrence is considered minimal.

Corrective Ac tion l

Valve NS03A was completely disassembled and inspected both visually and dimensionally. A 'new valve poppet and poppet pad was installed in the valve and a stellite overlay weld repair was made to the worn section of the valve

. body rib.

In addition, the valve stem and lantern ring were replaced due to indications of wear.

The cause of NS03A valve deterioration will be evaluated and corrective actions taken as deemed necessary.

Valve NS04A was repacked.

Af ter repairs were completed, both NS03A and NSO4A were retested satisfactorily.

The present acceptance criteria of 11.9 SCFH appears to be overly conservative for a Main Steam Isolation Valve and is currently being evaluated. A Technical Specification change request will be submitted, if the evaluation warrants 10 CFR 50, Appendix J does not require a leakage limit on individual o ne.

isolation valves.

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