ML19347E388
| ML19347E388 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/24/1980 |
| From: | Noell P, Stilwell T Franklin Research Ctr |
| To: | Polk P Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19347E387 | List: |
| References | |
| CON-NRC-03-79-118, TAC 12930, TAC 12931 TER-C5257-261-62-R01, NUDOCS 8104270210 | |
| Download: ML19347E388 (10) | |
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THIS REPORT SUPERSEDES ISSUE OF AUGUST 22, 1980
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TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSU RE ISOLATION VALVES WISCONSIN-MICHIGAN POWER COMPANY 1
POINT BEACH UNITS 1 AND 2 N RC DOCKET NO.
50-266, 50-301 NRC TAC NO. 1?930, 12931 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03-79-118 FRC TASK 261, 262
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Prepared by Franklin Research Center Author:
P. N. Noell The Parkway at Twentieth Street T. C. Stilwell Philadelpnia, PA 19103 FRC Group Leader:
? 5 50 ell Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:
?. J. Folk Octobei 24, 1980 This report was prepared as an account of work sponsored by an
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agency of the United States Government. Neither the United States l
Government nor any agency thereof, er any of their employees, j
makes any warranty, expressed or implied, or assumes any legal i
liability or responsibility for any third party's use, or the results of such use, of any information, apcaratus, product or process disclosed in this report, or recresents that its use by such third party would not infringe privately owned rights.
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i bb! Franklin Research Center 8104 M 0 g(O A Didsi n f The Franklin Institute a
We Beniarrun FranmLn Pamey PV.a Pa.151:3 (215)446-1000
1.0 INTRODUCTION
The NRC has determined that certain isolation valve configurations in systems connecting the high-pressure 7:imary Coolan System (PCS) to lower-pressure syste=s extending outside containmen are potentially significanc 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 eveces leading to the core melt is initiated' by the con-current. failure of two in-series check valves to function as a pressure isola-tion barrier between the high-pressure PCS and a lower-pressure system extend-ing beyond containment. This failure can cause an overpressurica: ion and rup-ture of the low-pressure sys:e=, resulting in a LOCA : hat bypasses containment.
The NRC has de: ermined 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 :es:ing, ul::asonic exa=ina: ion, or radiographic inspection. The NRC has established a program :o provide increased assurance such =ultiple isolatica barriers' are in place in all operating Ligh:
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Wa:e Reactor plan:s designated by DOR Generic !=plemen:atier Ac:ivi:y 3-45.
In a generic letter of Februzzy 23, 1980, the NRC requested all licensees
- o iden:ify :he following valve configurations which =ay exist in any of their f
plan: sys:e=s coc=unicating vi:h the PCS: 1) two check valves in series c: 2)
- vo check valves in series vich a motor-operated valve (MOV).
For plants in which valve configurations of concern are found to exist, licensees were further requested to indicate: 1) whether, to ensure integri:y j
of :he various pressure isola: ion check valves, continuous surveillance or i
periodic testing was cur:ently being condue:ed, 2) whe:her any check valvas of i
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concern vere known to lack integrity, and 3) whether plant procedures should i
i be revised or plant =odifica: ions be =ade to increase reliability.
Franklin Research Center (FRC) was requested by :he NRC :o provide tech-nical assis:ance to NRC's 3-4f activi:y by reviewing each licensee's sub=i::al l l
against criteria provided by the NRC and by 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 follev-ing five items =ust 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 contaicment;
- 4) the line must have one of the valve configurations shown in Figure 1; and
- 5) the pipe line =ust have a diameter greater than 1 inch.
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Figure 1.
Valve Configurations Designated by the NRC To 3e Included in This Technical Ivaluatice
2.2 Periodic Testing Criteria yor licensees whose plants have valve configurations of concern and choose to institute periodic valve leakage testing, the NRC has established criteria for frequency of testing, test conditions, and acceptable leakage rates.
These cri:eria may be summarised as follows:
2.2.1 Frequency of Testing Periodic hydros tatic leakage tes ting *. on each check valve shall be accom-
~~lished every time the plant is placed in the cold shutdown condition for p
refueling, each time the plant is placed in a cold shutdown condition fer 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been acco=plished in, the preceding 9 months,
each ti=e any check valve may have moved from the fully closed position (i.e., any time the differen-tial pressure across the salve is less than 100 psig), and prior to returning the valve to service af ter =aintenance, repair, or replacement work is performed.
2.2.2 Hydros:atic Pressure Criteria Leakage tes ts involving pressure dif ferentials lover than function pres-sure differentials are permitted in those types of valves in which service pressure will tend to di=inish :he overall leakage channel opening, as by pressing the disk in:o er ento :he seat wi:h grea:er force. Gate valves, check valves, and globe-type valves, having functica pressure differential applied over the seat, are exa=ples of valve applications satis fying this requirement.
k~ hen leakage :ests are made in such cases using pressures
!cuer than function maxi =u= pressure differen:ial, the observed leakage shall be adjus:ed to fune:icn =axi=u= pressure dif ferential value. This adjus:=ent shall be made by calcula:ica appropriate.co the test media and
- he ra:io between-tes: and fune: ion pressure dif feren:ial, assu=ing leak-age :o be direc:17 proper:ional :o the pressure differen:ial :o the one-half power.
2.2.3 Acceptable Leakage Ra:es:
Leakage rates less than or equal to 1.0 gpm are censidered accept-l able.
Leakage ra:es grea:er :han 1.0 gpm but less :han or ecual :o 5.0 e
l gp= are censidered acceptable if :he la:es: =easured rate has no:
i exceeded the ra:e de:er=ined by :he previous tes t by an amoun:
' o satis fy ALARA requirenen:s, leakage =ay be censured indiree:ly (as fr o=
- he perfor=ance of pressure inu.. -.s:s) if accomplished in accordance wi:h appreved precedures and supper:ed by co=putatices showing :ha: the me: hod 1
s capable of de=enstra:ing valve ec=pliance with the leakage criteria.
P00R OR B A
that reduces the margin between measured leakage rate and the maximum permissibic rate of 5.0 gpm by 50% or greater.
Leakage rates greater than 1.0 gpm but less than or equal to 5.0 e
gym are considered unacceptable if the latest measured rate ex-caeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
Leakage rates greater than 5.0 rpm are considered unacceptable.
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3.0 TECHNICAL EVALUATION
3.1 Licensee's Rasponse to the Generic Letter In response to the NRC's generic letter (Ref. l}, the Wisconsin-Michigan Power Company (WMP) enclosed a sketch in Reference 2 showing two systems which contain valve configurations of concern at Point Beach Units ?. and 2.
Ibese are in thz Residual Heat Removal System (low-pressure) and che cold leg of the Safety Injection System (intermediate-prassure).
The Licensee also stated that, "these valves (as per the enclosed sketch' will be verified closed by a combination of periodic testing and periodic ob s e rvation. "
It is FRC's understanding that, with WMP's concurrence, the NRC will di-rect WMP to change its Plant Technical Specifications as necessary to ensure that periodic leakage testing (or equivalent testing) is, conducted in accor-dance with the c.lteria of Section 2.2.
l 3.2 FRC Review of Licensee's Response FRC has reviewed the licensee's response against the plant-specific Piping l
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and Instrumentation Diagrams (P& ids) (Ref. 3] thst might have the valve con-figurations of concern.
- FRC has also reviewed the efficacy of instituting periodic testing for the check valves involved in this particular application with respect to the re-duction of the probability of an intersystem LOCA in the Residual Heat Re= oval
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and Safety Injection Syste= pipe lines.
In its review of the P& ids (Ref. 3} for the Point Beach Units 1 and 2, FRC found two following piping systems to be of concern:
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Ihe Residual Heat Recoval System, is composed of two sep. rate pip-ing lines (A and 3) connected directly to the reactor vessel, each line having two check valves and a motor-operated valve in one of the series configurations of concern.
The Safety Injection System, an intermediate-pressure system, is composed of cold and hot leg injection lines. The cold leg injec-tion piping is' connected to the loop A and 3 cold legs of the Primary Coolant System, while the hot leg injection piping, lines A and B, communicates with the reactor vessel directly. For both the hot and cold leg injection lines, the valve configuration of concern consists of two dheck valves and a motor-operated valve in series.,
It should be noted that a 10" cross-over line exists between Loop B, cold leg of the Safety Injection System, and the Residual Heat Removal. Sys tem.
This cross-over line, in conjunction with the loop B cold leg Safety Injection line, for=s a single check and motor-operated in series valve configuration of concern.
In each case the high-pressure / low-pressure interface is on the upstream side of the motor-operated valve (MOV). The two systems and their sopropriate valves are listed below for Units 1 and 2.
Residual Heat Removal System Line 1 high-pressure check valve, 853C high-pressure check valve, 853A high-pressure MOV, 852A, nor= ally closed (n.c.)
Line 2 high-pressure check valve, 853D high-pressure check valve 85:3 high-pressure MOV, 8523, n.e.
Safety Injection System Loop A, cold leg Line 1 high-pressure check valve, S67A
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l high-pressure check valve, 845A high-pressure MOV, 878D, n.o.
Line 2 high-pressure check valve, 867A high-pressure check valve, 845E high-pressure MOV, 878E, u.o.
loop B, cold' leg Line 1 high-pressure check valve, 867B high-pressure check valve, 845B high-pressure MOV, 878B, n.o.
Line 2 high-pressure check valve, 867B high-pressure check valve, 845F high-pressure MOV, 8787, n.o.
Reactor vessel, hot leg Line A high-pressure check valve, 653C high-pressure check valve, 845C high-pressure MOV, 878C, n.c.
Reactor vessel, hot leg Line B high-pressure check valve, 853D high-pressure check valve, 845D high-pressure MOV, 878A, n.c. -
Safety Injection (Cold Leg)-Residual Heat Removal Cross-Over Line high-pressure check valve, 867B high-pressure MOV, 720, n.c.
In accordance with the criteria of Section 2.0, FRC has found no other valve configurations of concern existing in this plant. These findings are
' consistent with the licensee's response (Raf. 2].
FRC reviewed the effectiveness of instituting periodic leakage testing of the check valves'in these lines as a means of reducing :he prob. ability of an intersystem LOCA occurring. FRC found that introducing a program of check valve leakage testing in accordance with the criteria summarized in Section 2.0 will be an efft:tive measure in substantially reducing the probability of an intersystem LOCA occurring in these lines, and a means of increasing the probability that these lines will be able to perform their safety-related functions.
It is also a step toward achieving a corresponding reduction in the plant probability of an intersystem LOCA in the Point Beach Units 1 and 2.
4.0 CONCLUSION
Point Beach Units 1 and 2 has been deter =ined to have valves in two of the configura: ions of concern (with two check valve and =otor-operated valve (MOV) in series configuration) for the Residual heat removal and the Safety Injec: ion syste=s, with a single check and MOV in series configuration 'for the RHR-SIS cross-over line.
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If 'n'MP modifies the Plan: Technical Specification for Point Beach Units 1 and 2 to incorporata periodic testing (as delineated in Section 2.2) for the check valves itemized in Table 1.0, then FRC considers this an acceptable means l
of achieving plant compliance with the NRC staff objec:ives of Raterence 1.
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Table 1.0 Primary Coolant System Pressure Isolation Valves System Check valve No.
Allowable Leakage
- Residual Huat Removal Line 1 853C 853A Line 2 853D 853B Safety Injection System Loop A, cold leg 867A 845A 845E Loop B, cold leg 8673 845B 845F R.V., hot leg line A 845C R.V., hot leg line B 845D 5.0 ' REFERENCES
[1]. Generic NRC letter, dated 2/23/80, from Mr. D. G. Eisenhut, De partment of Operating Reactors (DOR), to Mr. C. W. Fay, Wisconsin-Michigan Power Company (WMP).
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[2].
Wisconsin-Michigan Power Company's response to NRC's letter, dated 3/21/80,. from Mr. C. W. Fay (WMP) to Mr. D. G. Eisenhut (DOR).
[3]. List of examined P& ids:
FSAR Drawings of Point Besch Units 1 and 2:
Fig. 4.2-1 Fig. 6.2-1, Sh. 1 Fig. 6.2-1, Sh. 2
- To be provided by licensee at a future date in accordance with Sectica 2.2.3.
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Fig. 6. 2-1, Sh. 3 Fig. 9.2-1 i
Fig. 9.2-2 Fig. 9.2-4 Fig. 9.3-1 Fig. 9.3-2 Fig. 9.3-3 l
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