05000219/LER-1982-019, Forwards LER 82-019-/01T-0.Detailed Event Analysis Encl

Forwards LER 82-019-/01T-0.Detailed Event Analysis Encl

ML20071J620
Person / Time
Site:	Oyster Creek
Issue date:	04/15/1982
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20071J622	List: 05000219/LER-1982-019, Forwards LER 82-019-/01T-0.Detailed Event Analysis Encl
References
NUDOCS 8204270315
Download: ML20071J620 (3)

LER-2082-019, Forwards LER 82-019-/01T-0.Detailed Event Analysis Encl

Event date:
Report date:
2192082019R00 - NRC Website

text

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• GPU Nuclear f** . gg gf P.O. Box 388 Forked River, New Jersey 08731 609-693-6000 Writer's Direct Dial Number:

April 982 cp

\  % ;/g Mr. Ronald C. Ilaynes , Administrator g -

p Region I V '

U.S. Nuclear Regulatory Commission ,

# 2 U ,og? 7 ,2-631 Park Avenue  ; o

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King of Prussia, PA 19406 -

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Dear Mr. Ilaynes:

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Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 U \

Licensee Event Report Reportable Occurrence No. 50-219/82-19/OlT This letter forwards three copies of a Licensee Event Report to report Reportable Occurrence No. 50-219/82-19/0lT in compliance 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 Enforcement U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Director (3)

Office of Management Information and Program Control U.S. Nuclear Regulatory Commission l

Washington, D. C. 20555 i NRC Resident Inspector (1)

Oyster Creek Nuclear Generating Station Forked River, N. J. 08731 8204270315 g>

GPU Nuclear is a part of the General Pubhc Utihties System ,

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OYSTER CREEK NUCLEAR GENERATING S MTION Ebrked River, New Jersey 08731 Licensee Event Report Reportable Occurrence No. 50-219/82-19/0lT Report Date

. April 15, 1982 Occurrence Date March 31, 1982 Identification of Occurrence It was identified that an abnormal degradation of the primary contaiment existed, based on the results of leak rate testing performed on the Main Stem isolation drain valves V-1-106, V-1-107 and V-1-110, V-1-lll.'

M is event is considered to be a reportable occurrence as defined in the Techni-cal Specifications, paragraph 6.9.2.a'.3.

Conditions Prior to Occurrence

% e reactor was'in the cold shutdown condition at the time the occurrence was identified. Ibwever, the reactor was in various modes between the time period of January,1980 until the date of identification of the event.

Descriptien of Occurrence On February 8,1982, while performing local leak rate tests on Main Steam isola-tion drain valves V-1-106 and V-1-107 the results were found to be outside of the acceptable limit provided in the test procedure. On March 18,1982, the re-maining pair of Main Stem isolation drain valves, V-1-110 and V-1-lll, were leak tested, and the results of this test were also outside of acceptable limits.

Since V-1-110 and V-1-lll were tested as a pair, it could not be determined which valve was leaking. An additional test was performed and valve V-1-lll was deter-l mined to be the leaking valve.

t l- In the piping configuration containing these valves, a leak in either V-1-10E or V-1-107 and a leak in V-1-111 provides an abnormal flow path from the reactor vessel to the Condenser hotwells.

Apparent Cause of Occurrence 2e cause of the leakage was deterioration of valve internals.

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Licensee Event Report Page 2 Reportable Occurrence No. 50-219/82-19/0lT Analysis of Occurrence Isolation valves V-1-110 and V-1-lll were tested as a pair with a resulting total leakage past these valves in excess of 100 SCRI. (The actual total leak-age was not measurable due to the inability of the testing equipnent to measure leakage greater than 100 SCRI.) 'Ihe total leakage past isolation valves V-1-106 and V-1-107 when leak rate tested as a test pair was 49 Scal. 'Ihere-fore, the total leak past the four isolation valves in series or the primary contairment penetration was 49.0 SCRI.

If a IOCA had occurred during the event period, leakage fran this primary con-tairment penetration would be directed to the condenser "A" hotwell due to the piping configuration associated with these penetration isolation valves. At the condenser, primary contairment steam leakage would be condensed, mixed with, and contained in the condensate water system. Any primary contairment non-con-densible gas leakage would be contained in the condenser and off-gas systens.

Corrective Action Valve V-1-106 was repaired and passed the subsequent leak rate test. Valve V-1-lll was replaced and leak tested successfully. A further investigation in-to the suitability of these valves will be perforned and additional corrective action taken, if required.

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