ML19256B155

Forwards LER 78-111/03L-0

ML19256B155
Person / Time
Site:	FitzPatrick
Issue date:	01/19/1979
From:	Leonard J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19256B156	List: ML19256B155 ML19256B157
References
NUDOCS 7901240173
Download: ML19256B155 (1)

Text

,

POWER AUTHORITY OF THE STATE OF NEW YORK

- JAMES A. FITzPATRICK NUCLEAR POWER PLANT JOHN D. LEONARD, JR. P.O. BOX 41 Resident Manager Lycoming, New York 13093 January 19, 1979 33,3]334g JAFP-79-047 Mr. Boyce H. Grier United States Nuclear Regulatory Commission Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406

Reference:

Docket No. S0-333 Licensee Event Report: 78-111/03L-0

Dear Mr. Grier:

We have enclosed the referenced Licensee Event Report in accordance with Section 6.0 of Technical Specifications and USNRC Regulatory Guide 1.16.

If there are any questions concerning this report, please contact Mr. W. Verne Childs at 315-342-3840, Extension 207.

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Verytrulyyours[

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~ud N Jr. q,Y, h [hh,D.i, teonard JDL:VC:jjh Enclosure Re ident Manager CC: USNRC Director, Office of Inspection S Enforcement (30 copies)

USNRC Director, Office of Management information 4 Program Control (3 copies)

Internal Power Authority Distribution 790124 d / 7 3

ATTACllMENT TO LER 78-111/03L-0 Page 1 of 1 During routine plant startup operations, following a reactor scram and

. co'ntainment isolation, ECCS equipment was started automatically in response to a high drywell pressure signal. The high drywell pressure condition (approximately 2.7 psig) resulted because operations personnel did not properly return drywell pressure indication to its non-isolated condition and air (nitrogen) system leaks inside the drywell slowly increased pressure to the trip point. When the cause of the high drywell pressure condition was determined, the instrumentation valve lineup was corrected, the drywell pressure was restored to normal (approximately 1.7 psig) and ECCS equipment was restored to normal. Shortly after shutdown of RHR pumps B and D (10-P-3B and 3D) the "B RHR Loop Not Full" alarm came in.

Investigation of the "B RHR Loop Not Full" alarm revealed that the dis-charge check valve associated with the D RHR pump (10-P-3D) was stuck open or leaking. This allowed water in the B RHR Loop discharge header to drain back to the torus through the check valve and pump. Since a reactor cooldown was in progress and the A RHR Loop Pumps (10-P-3A and 3C) were operating in the torus cooling mode (See LER 78-109/03L-0) the operating staff made the decision to continue the cooldown and initiate repair of the check valve rather than to initiate the surveillance testing required by specification 4.5.A.3.a. This action satisfied the requirements of specification 3.5. A.6.

Approximately 45 minutes after initiation of the event, RHR Pump D was isolated to allow repair. Dissassembly of the check valve revealed that one (1) of two (2) valve disc retaining nuts had worked loose and was jammed between the valve seat and disc. The extra retaining nut was removed from the valve and the remaining retaining nut was pinned in accordance with vendor instructions.

Following reassembly of the valve, the pumps in the B RHR Loop were tested in accordance with Operations Surveillance Test F-ST-2B titled "RHR Pump Operability Test (ISI)" and returned to service in the standby mode approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the initiating event.

NOTE: LER 78-109/03L-0 is a related event.