ML19331D935

Primary Coolant Sys Pressure Isolation Valves,Brunswick Units 1 & 2, Technical Evaluation Rept

ML19331D935
Person / Time
Site:	Brunswick
Issue date:	07/21/1980
From:	Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Polk P Office of Nuclear Reactor Regulation
Shared Package
ML19331D934	List: ML19331D935
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-225, TER-C5257-225-9, NUDOCS 8009040315
Download: ML19331D935 (5)

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g TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES CAROLINA- POWER & LIGHT COMPANY BRUNSWICK UNITS 1~& 2 NRC DOCKET NO. 50-325, 50-324 NRC TAC NO. 12894, 12895 FRC PROJECT C5257 N RC CONTR ACT N O. NRC-03-79-118 FRC TASK 225,.229 Prepared by Franklin Research Center Author: P. N. Noell/T. C. Stilwel The Parkway at Twentieth Street Philadelphia, 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 I

liability or responsibility for any third party's use, or the results of I

such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

Oj _.[ Franklin Research Center A Division of The Franklin institute The Ben;amn FranMm Parn ey. Phila . Pa 19103 (215) 448 lCOO

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1.0 INTRODUCTION

The NRC staff Las determined that certain isolation valve configurations in systems connectin; the high pressure Primary Coolant System (PCS) to lower-pressure systems extending outside containment are potentially significant contributors to an intersystem losa-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 barrist 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 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 3-45.

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 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 3-45 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.

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 le ad 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.

Lt!d PCS7 I povl X I

= A Dk] A I

IEs

= Dk2 /

/

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t i/Iwovl I

kovl l HP: LP Figure 1. Valve Configurations Designated by NRC to be Included in This Te:hnical Evaluation

3.0 TECENICAL EVALUATION FRC h'as reviewed the response [Ref. 2} of the Carolina Power & Light Company (CPL) to NRC's generic letter [Ref.1] concerning the issue of PCS pressure isolation, valve configurations for Brunswick Steam Electric Plant Units 1 & 2.

. 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 CPL's findings. .

4.0 CONCLUSION

S In CPL's Brunswick Steam Electric Plant Units 1 & 2, 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. E. E. Utley, Carolina Power & Light Company (' CPL).

[2}. Carolina Power & Light Company's response to the generic NRC letter, dated 3/17/80, from Mr. E. E. Utley (CPL) to Mr. D. G. Eisenhut (DOR).

[3}. List of examined P& ids:

United Engineer & Construction Drawing:

9527-D-25016 (Rev. 7) Sh. 1 9527-D-25017 (Rev. 7) Sh. 2 9527-D-25nl8 (Rev. 10) Sh. 1 9527-D-25021 (Rev. 12) Sh. 1 9527-D-25022 (Rev. 9) Sh. 2 L

9527-D-25023 (Rev. 14) -

9527-D-25024 (Rev. 12) 9527-D-25025 (Rev. 12) Sh. 1 9527-D-25026 (Rev. 11) Sh. 2 9527-D-25027 (Rev. 10) Sh. 1 9527-D-25029 (Rev. 17) 9527-D-25037 (Rev. 9) 9527-D-25038 (Rev. 9) 9527-D-20041 (Rev. 17) 9527-D-25045 (Rev. 10) Sh. 3 9527-D-25047 (Rev. 10) 9527-D-25048 (Rev. 12) Sh. 2 9527-D-2274 (Rev. 5) 9527-D-26098 (Rev. 1) 9 4

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