ML18081B277
| ML18081B277 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/28/1980 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| IEB-79-06C, IEB-79-6C, NUDOCS 8004010447 | |
| Download: ML18081B277 (4) | |
Text
- o PS~G*
Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 March 28, 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Att:
Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management Gentlemen~
RESPONSE TO IE BULLETIN 79-06C NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 This letter is to advise you that our response to IE Bulletin 79-06C, dated August 29, 1979, for Salem 1, also applies to Salem 2.
A copy of the response is attached.
EAL:mw AS03 1 The Energy People Very truly yours, R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction ti
Frederick W. Schneider V1CP PrPsident Production August 29, 1979 Mr. Boyce H. Grier, Director U.S. Nuclear Regulatory Commission Off ice of Inspection and Enforcement Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Mr. Grier:
NRC IE BULLETIN NO.79-06C NO. 1 UNIT SALEM GENERATING STATION Pursuant to the subject bulletin, we hereby, submit the following response:
Short-Term Actions Item 1 A.
Station Emergency Procedures have been revised such that Reactor.Coolant Pumps are irrunedi~~ely tripped upon Reactor Trip with initiation of Safety Injection caused by low reactor coolant
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system pressure.
B.
A station operating memo has been issued re-quiring the presence of two licensed opera~ors Item 2 in the control room during operation in Modes 1, 2 and 3.
A series of Loss of Coolant Accident (LOCA) analyses for a range of break sizes and a range of time lapses between initiation of break and pump trip applicable to the 2, 3 and 4 loop plants has been performed by the Westinghouse Owners' Group.
A report summarizing the results of the analysis of delayed Reactor Coolant Pump trip during small loss of coolant accidents for West-inghouse NSSS will be submitted to Mr. D. F. Ross
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Boyce H. Grier, Director 8-29-79 Item 3 Item 4 by Mr. Cordell Reed on August 31, 1979.
In the report, maximum PCT's for each break size con-sidered and pump shutoff times have been provided.
The report concludes that' if the reactor coolant pumps are tripped prior to the reactor coolant system pressure reaching 1250 psia, the resulting peak clad temperatures are less than or equal to those reported in the FSAR.
In addition, it is shown that there is a finite range of break sizes and RCP trip times in all cases 10 minutes or later, which will result in PCT's in excess of 2200°F as calculated with conservative Appendix K models.
The operator in any event would have at least 10 minutes to trip the RCP's following a small break LOCA, especially in light of the con-servatisms in the calculations.
This is appropri-ate for manual rather than automatic action, based on the guidelines for termination of RCP operation presented in WCAP-9600.
The.Westinghouse Owners' Group has developed guidelines which were submitted to the NRC in Section 6 and Appendix A of WCAP 9600.
The analy-ses provided as the response to Item 2 are con-sistent with the guidelines in WCAP 9600.
No changes to these guidelines are needed for both LOCA arid non~LOCA transients.
The Owners' Group effort to revise emergency procedures covers many issues, including operation of the Reactor C9olant Pumps.
The action taken in response to item~l i~ sufficient as an interim measure in regards to these pumps.
The expected schedule for revising the LOCA, steamline break and steam generator tube rupture emergency pro-cedures is the following:
Mid-October:
Guidelines which have been reviewed by the NRC will be provided to each utility.
Appropriate utility personnel associated with writing procedures will meet with the Owners' Group Subcommittee on Procedures and Westinghouse to pro-vide the background for revising their emergency procedures.
1
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I Boyce H. Grier, Director 8-29-79 Item 5 1 to 2 months fromi Mid-October.
.1 to 4 months from:
Mid-Octobar Plant specific procedures will be revised.
Revised procedures will be imple-mented and operators trained.
Analyses related to inadequate core cooling and de-finition of conditions Wlder which a restart of "t.he RCP' s should be attempted will be. performa_d.
Resolution of the requirements for the analyses and an acceptable Bchedule for providing the
. analyses and guidelines and procedures result~_ng from the analyses will be arrived at between the Westinghouse Owners' Group and the NRC staff.
Lona Term Actions As discussed in our response to short-term item 2, we do not believe that automatic tripping of the RCP's is a required fWlction based on the analyses that have been performed and the guidelines that have bean developed for manual RCP tripping.
We propose that this item be discussed with t.he NRC staff following their review of the OWners' Group Submittal.
If you have any further questions on this matter we will be pleased to discuss t.'lem with you. Sine)
CC s Director, Office of In.spection and Enforcement USN RC Washington, DC 20555 Director, Office of Nuclear Reactor Regulation USN RC Washington, DC 20555
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