ML19331C920

Primary Coolant Sys Pressure Isolation Valves,Browns Ferry Units 1,2 & 3, Technical Evaluation Rept

ML19331C920
Person / Time
Site:	Browns Ferry
Issue date:	07/21/1980
From:	Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Polk P Office of Nuclear Reactor Regulation
Shared Package
ML19331C921	List: ML19331C920
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TAC-12892, TER-C5257-226, TER-C5257-226-7-8, NUDOCS 8008200121
Download: ML19331C920 (5)

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TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES

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TENNESSEE VALLEY AUTHORITY BROWNS FERRY UNITS L 2 & 3 NRC DOCKET NO. 50-259, 50-260, 50-296 NRC TAC NO. 12891, 12892, 12893 FRC PROJECT C5257 l N RC CONTRACT N O. NRC-03-79-118 FAC TASK 226 227 228 ,

Prepared by Franklin Research Center Author: P. N. Noell/T. C. Stilwell The Parkway at Twentieth Street Philadelnhia, PA 19103 FRC Group Leader: P. N. Noell Prepared for Nuclear Regulatory Commission Washington, 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 the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third l party would not inf ringe privately owned rights.

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l ILIU Franklin Research Center A Division of The Franklin Institute The Benjamin FraWn Parway. Pns.a . Pa 191 0 (215; 448-1000 8008200I21

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1.0 INTRODUCTION

The NRC staff has determined that certain isolation valve configurations in systems connecting the high pressure Primary Coolant System (PCS) to lower-pressare systems extending outside containment are potentially significant contributors to an intersystem Ioss-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 pressure 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 monitored, or if each valve is periodi-cally inspected by leakage testing, ultrasonic examination, or radiographic inspection. NRC has estdslished 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 Fe>ruary 23, 1980, the NRC requested all licensees to identify the following valve configurations which may exist in any of *. heir 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 j surveillance or periodic testing was currently being conducted, 2) whether any 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-nical assistance to NRC's B-45 activity by reviewing each licensee's submittal 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-ing five items must be fulfilled:

1) The high pressure system must 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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$ A l HP LP Figure 1. Valve Configura:icns Designa:ed by NRC to be Included in 2.is Technical Evaluation

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3.0 TECHNICAL EVALUATION

FRC has reviewed the response [Ref. 2} of the Tennesse Valley Authority (TVA) to NRC's generic letter [Ref.1] concerning the issue of PCS pressure isolation valva configurations for Brorns Ferry Nuclear Plant, Units 1, 2 & 3.

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 (P& ids) [Ref. 3} for piping systems that might have these valve configura-tions. In this review of the licensee's response against the P& ids and the identification criteria, FRC found no valve configurations of concern, thus verifying TVA's findings.

4.0 CONCLUSION

S In TVA's Browns Ferry Nuclear Plant, Units 1, 2 & 3, all piping systems larger than 1-inch diameter that are interconnected to the PCS are free of the valve configurations of concern. Therefore, no futher modifications to this plant's Technical Specifications are necessary on this account.

5.0 REFERENCES

[1]. Generic NRC letter, dated 2/23/80, from Mr. D. G. Eisenhut, Department of Operating Reactors (DOR), to Mr. L. M. Mills, Tennessee Valley Authority (TVA).

[2}. Tennessee Valley Authority response to the generic NRC letter, dated 3/14/80, from Mr. L. M. Mills (TVA) to Mr. D. G. Eisenhut (DOR).

[3}. List of examined P& ids:

General Electric Drawing:

728E900 Browns Ferry Nuclear Plant FSAR:

Figure 4.9-3 Tennessee Valley Authority Drawing:

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( hv. 15) 4 74811-1 4T4812-1 ( Rev . 14) 4 74813-1 (Rev. 11) 4T4814-1 (Rev. 15) 4 74817-1 (Rev. 13) 4T4854-1 (Rev. 5) e 4

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