ML20070L402
| ML20070L402 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/23/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070L399 | List: |
| References | |
| NUDOCS 8301030115 | |
| Download: ML20070L402 (3) | |
Text
r SAFETY EVALUATION Af1 Ell 0MENI NO.12 10 NPF-10 SAN ONOFHE t40 CLEAR GENERATING STATION, UNIT 2 00CKET NO.60-361 Introduction liy letters dated December 20, 21, and 22, 1982, Southern California Edison Company (SCE) on behalf of itself and the other licensees (San Diego Gas and Electric Company, the City of Anaheim and the City of Riverside), requested a revision to Technical Specification 3/4 1.2, Table 3.3-5.
The proposed change will delete the requirement that the corwent cooling water (CCLl) non-critical loop contain-nent isolation valves and ti t %'t critical /non-critical loop isolation valves isolate on low pressurizer p x'.sure.
All these valves will continue to isolate on high containment pressure ' Sout 3 psig). The CCll critical /non-critical loop isolation valves will c
.inue to isolate on low-low CCW surge tank level.
SCE states that renoval of t nssurizer Pressure-Low isolation signal from c
these valves will pernit tr:
.isten to continue cooling non-critical loop unp (RCP) motors and seals and the Control loads such as the Reactur.00, o
Elenent Drive Mechanism (CEDr ings during certain transient events.
Under the previous Technical Speci ic m s, these transient events would unnecessarily require that cooling to the P.
m tc and seal and CEDl1 winding be terminated.
SCE states that the continud r" allowed by the proposed change to the Technical Specification wis-in' unulative danage to the RCP punos seals caused by unnecessary wtr cro.< :a, of CCW. Minimization of such cumulative damage to the RCP seals will iM'co w J.ie 0,4iieni?ity of the RCPs and reduce the probability of RCP seal fai;ures.
Evaluation of the Proposed Channe As indicated by SCE in their letters, the proposed cho:.ge to the Technical Specifications will only affect the response of the CCW systen for those transient events which result in Pressurizer Pressure-Lov but not Containuent Pressure-High.
Such events are pressurizer pressure control systen failures, rain stean or feedwater system control systen or piping failures outside containment, and snall steam, feedwater and reactor coolant system piping failures inside containment. For these events, the following analyses ar.d conclusions are presented by SCE:
(1) The CCW Systen design has been reviewed and it has been verified that flow and heat capacity are adequate to sinultaneously serve all essential and non-essential loads with the exception of the Shutdown Cooling Heat Exchangers (SDCHX).
The SDCHX's are isolated until they receive the Containnent Spray Actuation Signal at approxinately 16 psig containnent pressure. Because the proposed signal to isolate the essential from the non-essential loads occurs at approxinately 3 psig containnent pressure (on Containment Isolation Actuation Signal), the CCW systen capacity will not he exceeded.
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Isolation of the critical CCH loops f rom the non-critical 1000 occurs on low-low CCW surge tank level, thereby protecting the critical loops from non-critical loop failures.
SCE states that each of the safety analyses of Section 15 of the FSAR has been examined, and the proposed chanc,e will not inpact the analyses.
(3) While the proposed change will reduce the diversity of the actuation signal to the non-critical loop CCW containment isolation valves and the critical /
non-critical CCW isolation valves, the overall result of the change will be to allow operation of the RCPs during a wider range of transient events.
Since RCP operation can mitigate the consequences of nany accident sequences, SCE argues that the proposed change results in increased plant safety, on balance.
The FRC staff has reviewed the SCE letters, and has discussed the issue with SCE in a meetint) on December 21, 1982.
The staff concurs with SCE's conclusions, for the reasons given above. Consequently, the staff finds the proposed changes to Technical Specification.',/a.3.2 to be acceptable.
Evaluation of Related Issues During the course of the staff's review of the proposed change, it becane clear that the non-critical CCW containment penetr.itions do not meet the a;>plicable staff criteria or the criteria defined in the FSAR. Specifically.
the non-critical CCU loop can not be shown to meet the criteria specified in the FSAR for systens which neet General Design Criterion (GDC) 57, since the 1000 inside containment is not nissile and pipe-whip protected, and the components served are not seismic Category I.
Therefore, the isolation provisions nust neet GDC 56, which requires two autonatic isolation valves for each line penetrating containment.
The present design has two isolation valves per line, but only one of the valves isolates automatically. The other has remote-manual actuation.
At a meeting on December 22, 1982, this issue was discussed with SCE. By its letter dated December 22, 1982, SCE comitted to correct the situation within 90 days by changing the renote-nanual valves (liv-6223 and 6236) to automatic isslation. The 's0 day time period was justified on the basis of naterial delivery time and installation and testing time.
In the interin, justification for continued operation will be based on procedures requiring operator verification of non-critical 1000 isolation, and operator action to close the renote-nanual i
valves should the automatic valves fail to close.
SCE further stated that the renote-manual valves will be unlocked to allow closure by the operators, rather thar, locked open as stated in the FSAR.
The NRC staff has reviewed the SCE letter of December 22, 1982, and has discussed the issue with SCE, and has concluded that operation prior to installation of the automatic isolation signal to valves HV-6223 and 6236 is acceptable, l
based on (1) autonatic isolation of valves HV-6211 and 6416 which are in series with HV-6223 and 6236, and (2) the ability to pronptly isolate valves HV-6223 and I
6236 from the control room should the need arise.
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3-j Environuental Consideration The tfRC staff has deternined that this amendment does not authorize a change in effluent types or total amount nor an increase in power level and will not result in any significant environnental inpact.
Ifaving nade this determination, we have further concluded that this anendnent involves action which is insignificant fron the standpoint of environmental impact and pursuant 10 CFR Section 51.5(d)(4), that an environmental impact statenent or negative declaration and environnental impact appraisal need not be prepared in connection with the issuance of this anendnent.
Conclusion Based upon our evaluation of the proposed changes to the San Onofre, Unit 2 Technical Specifications, we have concluded that:
(1)becausethisamendment does not involve a significant increase in the probability or consequences of accidents previously considered, does not create the possiblity of an accident of a type different from any evaluated previously, and does not involve a significant decrease in a safety margin, this anendment does not involve a significant safety hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed nanner, and (3) such activities will be con-ducted in compliance with the Cornission's regulations and the issuance of this amendrent will not be inimical to the connon defense and security or to the health and safety of public.
We, therefore, conclude that the proposed change is acceptable.
Dated: DEC 2 31982 h
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