ML19343C612
| ML19343C612 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/24/1980 |
| From: | Noell P, Stilwell T FRANKLIN INSTITUTE |
| To: | Polk P Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19343C599 | List: |
| References | |
| CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-118, TER-C5257-11B, NUDOCS 8103240605 | |
| Download: ML19343C612 (8) | |
Text
ATTACHMENT 1 1
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TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES CONNECTICUT YANKEE ATOMIC F0WER COMPANY HADDAM NECK UNIT 1 NRC CCCKET No. 50-213 N AC TAC NO.
12915 FAC P ACJECT C5257 N AC CONTAACT NC. N AC-03-73-118 FAC TASK 219 Prepared cy Franklin Research Center Autncr: ?. N. Scell The Parkway at Twentieth Street T. C. Stil. ell Pniladelphia, RA 19103 FRC Grouc Leader:
?. N. Neell Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: ?. J. Pelk Oc cber 24, 1980 This report was prepared as an account of acrk sconscred cy an agency of the United States Government. Neitherine United States Government nor any agency tnereef, er any of their empicyees.
mtxes any warranty, expressec or implied. cr assumes any legal liabi'ity or responsibility for any third party's use, or the results of sucn use, of any information, acparatus, crocuct cr precess cisc!csed in this rescrt, or represents that its use by such thirc party would net infringe privately owned rignts.
A 000_I Franklin Research Center A DMsien of The Frankin insatute The sen % rr nun ;.m
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1.0 INTRODUCTION
The NRC has determined that certain isolation vaive configurations in systems 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 con-failure of two in-series check valves to function as a pressure isola-current tion barrier between the high pressure PCS and a lower pressure system extend-ing beyond containment. This failure can cause an overpressurication and rup-cure of the low pressure system, resulting in a LOCA that bypasses containment.
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 te;cing, ultrasonic examination, or radiographic inspection. The NRC has estaalished a program to provide increased assurance that such multiple isolation barriers are in place in all operating Light Water Reactor plants designated by 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 are found to exist, licensees were further requested to indicate: 1) whether, to ensure integrity of the various pressure isolation check valves, continuous surveillance or periodic testing was currently being conducted, 2) whether any check valves of concern were known to lack integrity, and 3) whether plant procedures 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 3-43 activity by reviewing esc's licensee's summittal.
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 CRITT.RIA 2.1 Identification Criteria For a piping system to have a valve configura:ica of concern, the follow-ing five i: ems mus: he fulfilled:
- 1) The high pressure system must be connected to the Pri=ary Coolan:
System;
- 2) there must be a high pressure / low pressure in:erface present in the line;
- 3) -dis same piping must eventually lead outside containment;
- 4) the line cmst have one of :he valve configura: ions shewn in Figure 1; and
- 5) the pipe line mus: have a dianecer greater than 1 inch.
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- LP Figure 1.
Valve Configurations Designated by he NRC To 3e Included in his Techni:a1 Evalua:icn
2.2 Periodic Test.ng Cri:eria For licensees whose plants have valve configurations of concern and choose to institute periodic valve leakage esting, the NRC has es:ablished cri:eria for frequency of testing, ces: conditions, and accep:able leakage rates.
These criteria may be summarized as follcws:
2.2.1 Frequency of Tes ing Periodic hydros ta:ic leakage ces:ing* on each check valve shall be accom-plished every :ime the plane is placed in the cold shu:dewn condition for refueling, each time :he plan: is placed in a cold shu:down conditica for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has no: been acccmplished in.:he preceding 9 mon:hs, each time any check valve may have moved from :he fully closed posicion (i.e., any time the differen-tial pressure across the valve is less than 100 psig), and prior to returning :he valve :o service af:er maintenance, repair, or replacement work is performed.
2.2.2 Hydrosta:ic pressure Cri:eria Leakage ces:s involving pressure differencials lower than fune: ion pres-sure differentials are permi::ed in : hose types of valves in which service pressure will :end :o diminish :he overall leakage channel opening, as by pressing :he disk in:o or en:o :he sea: wi:h grea:er force. Gate valves, check valves, and globe-:ype valves, having fune:ica pressure differen:ial applied over :he seat, are examples of valve aoplications satis fying this requiremen:._ When leakage :es:s -are made in such cases using pressures lower than fune: ion maximum pressure differencial, :he cbserved leakage shall be adjus:ed to function maxi =um pressure dif feren:ial value. This adjus: ment shall be made by calculation appropria:e :o :he :es: media and
- he ra:io between tes: and fune: ion pressure differential, assuming leak-age :o be directly propor:ional :o :he pressure differential :o :he one-half power.
2.2.3.Accep:able Leakage Ra:es:
Leakage races less chan or equal :o.l.0 gpm are considered accep:-
i e
able.-
Leakage races grea:er :han 1.0 -gpm but less :han or equal to 5 0 e
gpm are considered -acceptable if the la:es: measured race has no:
t exceeded the rate determined by the previous tes: by an amoun:
- To satis fy ALARA requirements, leakage may be measured indirec:1y (as Erom the performance of pressure indica: ors) if accomplished in accordance wi:h approved procedures and supported by computa: ions showing tha: the me: hod is capable of demonstra:ing valve compliance wi:h :he leakage cri:eria.
L 3
s
J that reduces the me 3 n between the measured leakage rate and the i
maximum permissible 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
gpm are considered unacceptable i' the latest measured rate ex-ceeded 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 gpm are considered unacceptable.
e
3.0 TECHNICAL EVALUATION
3.1 Licensee's Response to tb e Generic Letter
.In responi7 to the NRC's generic letter (Ref.1}, the Concecticut Yankee Atomic Power Company (CYA) stated (Ref. 2} that, "CYA?CO has reviewed all piping systems, including the low pressure safety injection system, which communicate with :ne reactor coolant syste= (RCS), penetrate con:ainment, and derate to a lower pressure system, and has determined : hat no event V valve configurations, as described in Reference (1), exist at the Haddam Neck Plant."
Since the licensee identified no valve ufigurations of concern, no sur-veillance techniques were reported for any valves. Nevertheless, valve con-figurations of concern were later discovered by FRC.
It is FRC's understanding that, with CYA's concurrence, the NRC will direct CYA to change its Plant Technical Specifications as necessary to ensure that periodic leakage testing (or equivalent testing)-is conducted in accor-dance wi:5 the criteria of Section 2.2.
3.2 FRC Review of Licen. ae's Response FRC-bas reviewed the licensee's response against the plant-specific Piping and Instrumentation Diagrams (PSIDs) [Ref. 3 } that might have the valve con-figurations of concern.
FRC has also reviewed the efficacy of instituting periodic. tecting for the check valves involved nin this particular application with respect to the re-duction of the probability of an intersystem LOCA in the High-Pressure Safety '
Injection and Core Deluge System piping' lines.
4
In its review of the P& ids [Ref. 3] for Haddam Neck Unit 1. FRC found the two following piping systems to be of concern:
The valve configuration of concern existing in each of the four cold-leg branches of the "gh-Pressure Safety Injection System con-sists of a single motor-ope sted valve (MOV) in-series with a single check valve with the MOV positioned closer to the reactor vessel than the check valve. The high pressure / low-pressure interface exists on the upstream side of the single check valve.
The Core Deluge System, composed of two 6-in branches leading to four separate reactor vessel head lines, has the sa=e valve configu-ration of concern as described for the High-Pressure Safety Injec-tion System.
The valve configurations for both systems are itemized below:
4 High-Pressure Safety Injec.
- n system Looo 1, cold lez h.gh pressure MOV, SI-MOV-361A, normally open (n.o.)
righ pressure check valve, SI-CV-362A Loon 2, cold leg high pressure MOV, SI-MOV-3613, n.o.
high pressure check valve, SI-CV-S623 Loco 3, cold leg high-pressure MOV, SI-MOV-361C, n.o.
high pressure check valve, SI-CV-362C Leon 4, cold leg high pressure MOV, SI-MOV-361D, n.o.
high pressure check valve, SI-CV-3623 Core Deluge System 3 ranch A high pressure MOV, CD-M07-371A, n.o.
high-pressure check valve, CD-CV-372A, n.o.
Branch B high pressure MOV, CD-MOV-371B, n.o.
high-pressure check valve, CD-CV-3723 In accordance with the criteria of Section 2.0, FRC found no other valve configurations of concern existing in this plant.
FRG reviewed the ef fectiveness of instituting periodic leakage testing of the check valves ta these lines as a means of reducing the probability of an intersystem LOCA occurring. FRC found that introducing a program of check valve leakage testing in accordance with the :riteria summarized in Section 2.0 will'be an ef fective measure in subt tanttally 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 planc probability of an intersysten LOCA in 9addam Neck Unit 1.
4.0 CONCLUSION
3ased on the 'previously docketed information and drawings made available for FRC review, FRC found that the High-Pressure Safety Injection and Core Deluge Systems in Haddam Neck Unit 1 contain a valve configuration of concern (identified in Figure 1).
Thus, if the licensee's reviev of the valving con-figuration in the High-Pressure Safety Injection and Core Delgue Systems con-firms FRC's finding,. then the valve configurations of concern existing in Haddam Neck Unit 1. incorporate the valves listed in Table 1.0.
If CYA modifies the plant Technical Specifications for Haddam Neck Unit 1 to incorporate periodic testing (as delineated in Section 2.2) for the check
. valves itemised in Table ~1.0, then FRC considers this an acceptable means of achieving plant compliance with' the NRC staff objectives of Reference 1.
.s.
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Table 1.0 Primary Coolant System Pressure Isolation Valves System Ch eck Valve No.
Allowable Leakage
- High-Pressure Safety Injection Loop 1, cold leg SI-CV-36 2A Loop 2. cold leg SI-CV-3623 Loop 3, cold leg SI-CV-362C JLoop 4, cold leg SI-CV-3 6 2D Core Deluge Branch A CD-CV-372A 3 ranch 3 CD-CV-3 7 23
5.0 REFERENCES
1.
Generic NRC letter, dated 2/23/30, from Mr. D. G. Eisenhut, Department of Operating Reactors (DOR), to Mr. W. G. counsil, Connecticut Yankee Atomic
. Power Company (CYA).
2.
Connecticut Yankee Atomic Power Company's response to NRC's letter, dated 3/13/30, from Mr. W. G. Counsit (CYA) to Mr. D. G. Eisenhut (DOR).
3.
List of examined P& ids:
Westinghouse Drawings of Haddam Neck Unit 1:
540 F 415 (Rev. 11) 540 F 416 (Rev. 3) 540 F 417 (Rev. 9) 540 F 419 (Rev. 9)-
540 F 420 (Rev. 10) 647 J 282 (Rev. 9) 2980 44-F (Rev. 7)
- To ~be' provided _by licensee at a future date in accordance with Section 2.2.3.
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