ML19344D913

Primary Coolant Sys Pressure Isolation Valves,Oyster Creek Unit 1, Technical Evaluation Rept

ML19344D913
Person / Time
Site:	Oyster Creek
Issue date:	07/21/1980
From:	Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Polk P Office of Nuclear Reactor Regulation
Shared Package
ML19344D908	List: ML19344D913
References
CON-NRC-03-79-118, CON-NRC-3-79-118, TASK-06-04, TASK-6-4, TASK-RR TER-C5257-252, NUDOCS 8008260329
Download: ML19344D913 (4)

Text

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O TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES l JERSEY CENTRAL POWER & LIGHT COMPANY  !

OYSTER CREEK UNIT 1 NRC DOCKET NO. 50-219 NRC TAC NO. 12919 FRC PROJECT C5257 NRC CONTR ACT NO. NRC-03-79-118 FRC TASK 252 l

Prepared by Franklin Research Center Author: P. N. Noell/T. C. S tilwell '

The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader. P. N. Noell Prepared for Nuclear Regulatory Commission Washin0 ton, D.C. 20555 Lead NRC Engineer: P. J. Polk July 21, 1980 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or tne results of

! such use, of any information, apparatus, product or prc.ccas

! disclosed in this report, or represents that its use oy such third party would not infringe privately owned rights.

b ._ Franklin Research Center A Division of The Frankhn Insttute TFe Ben.arn.n l 'amo.ri F a a.a, Pva in 1(*103121* ; 44e t r(A 8008260 32$ - .-- . _ -_ __ . . _ _

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1.0 INTRODUCTION

The NRC staf f has determined that certain isolation valve configurations in systcms connecting the high pressure Primary Coolant System (PCS) to lower-pressure systems extending outside containment are potentially significant contributors to an intersystem loss-of-coolant accident (LOCA). Such configu-rations have been found to represent a significant factor in the risk computed for core melt accidents. The sequence of events leading to the core melt is initiated by the failure of two in-series check valves to function as a pres-sure isolation barrier between the high pressure PCS and a lower pressure system extending beyond containment. This causes an overpressurization and rupture of the low presso-r system, which results in a LOCA that bypasses con-tainment.

The NRC has determined that the probability of failure of these check valves as a pressure isolation barrier can be significantly reduced, if the pressure at each valve is continuously mv.acored, or if each valve is periodi-cally inspected by leakage testing, u!' .;onic examination, or radiographic inspection. NRC has established a program to provide increased assurance that such multiple isolation barriers are in place in all operating Light Water Reactor plants designated DOR Generic Implementation Activity B-45.

In a generic letter of February 23, 1980, the NRC requested all licensees to identify the following valve configurations which may exist in any of their plant systems communicating with the PCS: 1) two check valves in series or 2) two check valves in series with a motor-operated valve (MOV). For plants in which valve configurations of concern were found to exist, licensees were further requested to indicate: 1) whether, to ensure integrity, continuous surveillance or periodic testing was currently being conducted, 2) whether any i valves of concern were known to lack integrity, and 3) whether plant proce-dures should be revised or plant modifications be made to increase reliability.

Franklin Research Center (FRC) was requested by the NRC to provide tech-j nical assistance to NRC's B-45 activity by reviewing each licensee's submittal l against criteria provided by the NRC and verifying the licensee's reported findings from plant system drawings. This report documents FRC's technical review.

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2.0 CRITERIA 2.1 Identification Criteria For a piping system to have a valve configuration of concern, the follow-inr, five items must be fulfilled:

1) The high pressure system cust be connected to the Primary Coolant System;

2) there must be a high pressure / low pressure interface present in the line;

3) this same piping must eventually lead outside containment;

4) the line must have one of the valve configurations shown in Figure 1; and

5) the pipe line must have a diameter greater than 1-inch.

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$ HPm N  : LP Figure 1. Valve Cenfigura:iens Desipated by NRC te be Included in This Technical Evaluation

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3.0 TECHNICAL EVALUATION

FRC has reviewed the response [Ref. 2] of the Jersey Central Power and Light Company (JCP) to NRC's generic letter [Ref.1] concerning the issue of PCS pressure isolation valve configurations for Oyster Creek Plant.

The licensee stated that none of their piping systems have any of the valve configurations of concern, as described by the identification criteria.

FRC independently checked the plant Piping and Instrumentation Diagrams (P6 ids) [Ref. 3] for piping systems that might have these valve configura-tions. In this review of the licensee's response against the P6 ids and the identification criteria, FRC found no valve configurations of concern, thus verifying JCP's findings.

4.0 CONCLUSION

S In Jr. s 0 ster Creek Plant, all piping systems larger than 1-inch dia-meter that are interconnected to the PCS are free of the valve configurations of concern. Thersfore, no futher modifications to this plant's Technical Specifications ate necessary on this account.

5.0 REFERENCE S (1). Generic NRC letter, dated 2/23/86, from Mr. D. G. Eisenhut , Department of Operating Reactors (DOR), to Mr. I. R. Fidarock, Jr. , Jersey Central Power and Light Co=pany (JCP).

[2]. Jersey Central Power and Light Company's respo:'se to the generic NRC letter, dated 3/17/80, from Mr. I. R. Finbrock (JCP) to Mr. D. G.

Eisenhut (DDR).

[3). List o f examined P61Ds:

General Electric Drawing:

1*SP711 (Rev. 7) 237E726 (Rev. 10) 14SF444 (Rev. 11) 237E798 (Rev. 10) 14SF723 (Rev. 3) 2406 (Rev. 8) 197ES71 (Rev. 7) 706E249 (Rev. 3) 237E457 (Rev. 15) S85D7S1 (Rev. 11) 886D403 (Rev. 2)