05000287/LER-1981-005, Forwards LER 81-005/03L-0.Detailed Event Analysis Encl

Forwards LER 81-005/03L-0.Detailed Event Analysis Encl

ML20063E904
Person / Time
Site:	Oconee
Issue date:	04/22/1981
From:	Parker W DUKE POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20063E907	List: 05000287/LER-1981-005, Forwards LER 81-005/03L-0.Detailed Event Analysis Encl
References
NUDOCS 8208310076
Download: ML20063E904 (3)

LER-2081-005, Forwards LER 81-005/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2872081005R00 - NRC Website

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/ ice Poggictaev ICL E*=O'e E: & # t 4 704 sec.- a coucwo April 22, 1981 va-ess Mr. James P. O'Reilly, Director U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Re: Oconee Nuclear Station Docket No. 50-287

Dear Mr. O'Reilly:

Please find attached Reportable Occurrence Report R0-287/81-05 This report is submitted pursuant to*0conee Nuclear Station Technical Specification 6.6.2.1.b(2),

which concerns operation in a degraded mode permitted by a limiting condition for operation, and describes an incident which is considered to be of no signi-ficance with respect to its effect on the health and safety of the public.

V y truly yours,

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/N%. LD Icw y .'

William O. Parker, Jr./

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JLJ:pw Attachment cc: Director Mr. Bill Lavallee Office of Management and Program Analysis Nuclear Safety Analysis Center U.S. Nuclear Regulatory Commission Post Office Box 10412 1 Washington, D. C. 20555 Palo Alto, California 94303 l

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8208310076 8:0422 l PDR ADOCK 05000287 S PDR

. . , . o DUKE POWER COMPANY OCONEE UNIT 3 Report Number: R0-287/81-05 Report Date: April 22, 1981 Occurrence Date: March 23, 1981 Facility: Oconee Unit 3, Seneca, South Carolina Identification of Occurrence: Failure of Engineering Safeguard Valve 3CS-5 Conditions Prior to Occurrence: 100% FP Description of Occurrence: At 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on March 23, 1981, engineering safe-guard valve 3CS-5 failed to operate. While attempting to open valve 3CS-5 during testing, the control switch indicator indicated the valve to be only partially open. Valve 3CS-6 was opened in orr.er to drain the quench tank.

Water was pumped from the tank at a slow rate until a low level was obtained.

Valve 3CS-6 was then failed shut to isolate valve 3CS-5. This constitutes operation in a degraded mode per Technical Specification 3.6.3.b(2) and is thus reportable pursuant to Technical Specifica tion 6.6. 2.1.b(2) .

A decision was made to repair valve 3CS-5 after a Unit 3 turbine trip on April 9, 1981. Personnel entered the Reactor Building to inspect the valve and found the following:

1. The valve was in the intermediate position (neither full-open or full-closed).

2. The Limitorque valve operator would not stroke the valve, but the valve could be operated manually.

3. Personnel checked the limit switches on the operator and found them to be properly set. It was reported that the valve or operator was binding, causing the torque limit switches to trip and thus, the i

valve not to cycle.

Apparent Cause of Occurrence: Since the valve could be operated manually, it was determined that the valve operator was the primary cause of the problem.

Since time was not available to complete the repair, the decision was made to maintain administrative control over the valve during continued operation.

Analysis of Occurrence: In the event of a LOCA or initiation of an Engineering Safeguard System, Reactor Building integrity will be maintained by valve 3CS-6.

During the opening of valve 3CS-6, administrative controls are maintained to insure capability to isolate the valve if necessary. Hence, the health and safety of the public were not adversely affected by this incident.

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Corrective Action: Valve 3CS-6 has been failed shut, thus isolating valve 3CS-5. On April 9, 1981, an investigation was made on valve 3CS-5, and the problem was determined to be in the operator. During the time 3CS-6 is open, administrative control is maintained to insure the capability to shut 3CS-6 if necessary.

During the'next Unit 3 outage, this valve / operator will be inspected and disassembled to determine why it failed and what further corrective actions are needed. While the valve / operator is being repaired, the valve will be preventive maintenanced and the diaphragm replaced. Once repaired, testing of 3CS-5 will resume. If it becomes evident that the valve must be repaired before the next Unit 3 outage, steps will be taken to do so.

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