ML25196A485

03-14-75 General Electric Standard Safety Analysis Report (GESSAR-238)

ML25196A485
Person / Time
Issue date:	03/14/1975
From:	Kerr W Advisory Committee on Reactor Safeguards
To:	Anders W NRC/Chairman
References
Download: ML25196A485 (1)

Text

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 14, 1975 Honorable llilliam A. Anders Chairman U.S. Nuclear Regulatory Commission Washington, D. C.

20555

Subject:

GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR-238)

Dear Mr. Anders:

At its 179th Meeting, March 6-8, 1975, the Advisory Com."'!l:l.ttee on Reactor Safeguards completed a review of the General Electric Standard Safety Analysis Report (GESSAR).

GESSAR-238 provides the safety information for a reference system consisting of a single BWR-6/1-iark III nuclear system, with a rated core th"':-:nal power of 3579 MW(t), and of the associated systems including t'!e reactor bu,Hding (the shield building and containment), fuel building, auxiliary building, diesel generator buildings, control building, radwaste building,, and the off-gas system.

Subcommittee meetings were held with representatives of the General Electric Company and the Nuclear Regulatory Coramission (NRC) Staff on July 1, 1974, and September 11, 1974, in Washington, D. C., on November 9, 1974, in Bloomington, Minnesota, and on January 18, 1975, in Washington, D.C.

The Committee also had the benefit of the documents listed below.

Site envelope parameters are included in GESSAR and application of GESSAR will require that specific site evaluations be made to confirm the acceptability of ~he site within the GESSAR design.

'rhe use of GESSAR for multiple reactor units at a single station will also require review of the safety-related components of plant duplication and layout.

Safety-related interfaces between the reference system and the balance of plant are specified in GESSAR.

Since the utility-applicant is responsible for instituting the qu,1lity assurance programs necessary to assure that all safety-related interfaces have been identified and that all safety-related requirements ar.-1 being fulfilled, the Committee will review these matters in more detaLi.. with the Applicants on a case-by-case basis. The Committee recom.~ends that, during the design, procure-ment, construction and startup, timely and appropriate interdisciplinary systea analyses be carried out to assure complete functional compati-bility across each interface for an en~ire spectrum of anticipated operations and postulated design basis accident conditions.

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Honorable William The NRC Staff has identified 13 items requiring resolution prior to issuing their Preli~inary Design Approval (PDA).

The Com.~ittee believes that all of these matters should be resolved in a manner satisfactory to the NRC Staff. The Committee wishes to be kept informed regarding the resolution of the following items:

1.

Seismic capability of the offgas system.

2.

Provisions to satisfy the single-failure criterion for the RHR system.

3. Additional requirements to be imposed if continuous venting of the containment is used.

4.

Evaluation of the performance of the emergency core cooling systems using evaluation models meeting the requ.!.:.*canents of 10 CFR 50.46, Appendix K.

The latest ACRS reports on nuclear generating stations utilizing the BWR-6/M.ark III systems were the December 12, 1974 reports on the Allens Creek Nuclear Generating Station, Units 1 and 2, and the Perry Nuclear Power Plant, Units 1 and 2.

In these reports, the ACRS has reco1lli.~ended that the ongoing R&D programs be used to fully resolve issues involving the Mark III containment design prior to completion of the affected portions of the plant. Further, additional generic matters, which include anticipated transients without scre.m (ATWS) and possible pump overspeed during a loss of coolant accident, should be dealt with appropriately by the NRC Staff. It is expected, that these items will be resolved in a manner satisfactory to the NE.C-Staff.following Preliminary Design Approval (PDA) of GESSAR and prior to Final Design Approval (FDA).

During this interim period, the COllllD.ittee will continue to revie\\/ these items on a case-by-case basis as well as/through other appropriate ACRS Subcommittee meetings and full Committee meetings.

The Committee has not reviewed modifications which are expected to be made in the mm./6 8x8 fuel.

Such modifications and any other proposed changes will be reviewed when the appropriate documentation has been submitted and the improvements sought can be evaluated.

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Honorable William The introduction of new features in the instrumentation and control systems has been submitted through the specification of functional designs and design criteria which the NRC Staff has found to be adequate for the PDA.

As in previous reports on related matters the Committee recommends that the NRC Staff determine the necessary environmental and reliability tests, including in situ tests where desirable for qualifi-cation of the new systems.

In another matter relating to a periodic testing provision, the General Electric Company has committed to a study of the improvement of the testability of the automatic depressurization system.

On all these issues involving instrumentation and control, the Committee will use the case-by-case basis to ascertain progress of the work until the GESSAR design has progressed to the stage where Final Design Approval is achieved.

The Committee will need to review the development and proof testing of the fast scram system, and the implementation of the proposed Reactor Hanual Control System along with the provisions for ganged rod with-drawal.

The Committee believes that the General Electric Company and the NRC Staff should continue to review GESSAR for design changes that would further improve industrial security features.

'fhe GESSAR design should include provisions which anticipate the maintenance, inspection, and operational needs of the plant throughout its service life, including cleaning and decontamination of the primary coolant system, and eventual decomnissioning.

In particular, the Committee believes that the NRC Staff and the General Electric Conpany should review raethods and procedures for reraoving accumulations of radioactive.contamination whereby maintenance and inspection programs can be more effectively and safely carried out.

The Committee believes that methods that seek to develop reference systems through standardization and through replication need to be coupled with ongoing programs that will permit changes which improve safety and which, when justified, would be bplemented in a timely manner.

Use of reference systems should lead to more efficient and effective licensing reviews.

Programs such as GESSAR will contribute to this process.

A transition period will be required in which the Committee would still give considerable attention to the items noted, on a case-by-case basis.

The Committee believes that, subject to the above comments and to succnssful completion of the R&D prgrams, GESSAR-238 can be success-fully engineered to serve as a reference system.

References attached, Sincerely yours,

~

William Kerr Chairman 2197

Honorable William References

1.

BWR/6 Standard Safety Analysis Report, Volume 1 through 7.

2.

.AJJ.endraents 1 through 28 to the Standard Safety Analysis Report.

3.

General Electric Company letters and reports:

a.

July 31, 1973 letter forwarding proprietary information in support of the information made public in the safety analysis report.

b.

August 31, 1973 letter forwarding proprietary f,1el data.

c.

September 28, 1973 letter forwarding proprietary information regarding core power distribution.

d.

December 28, 1973 letter regarding interfaces and electrical systems.

e.

November 6, 1974 letter regarding physics verification and number of safety/relief valves.

f.

February 19, 1974 letter regarding ATt1S.

4.

AEC/NRC Staff letters and reports:

a.

October 11, 1974 draft Safety Evaluation Report.

b.

November 13, 1974 Safety Evaluation Report.

c.

December 7, 1974 Supplement No. 1 to the Safety Report.

Evaluaion

d.

January 30, 1975 letter regarding reevaluation of the high pressure drywell test.

e.

February 21, 1975 Supplement No. 2 to the Safety Evaluation Report.

f.

March 4, 1975 Supplement No. 3 to the Safety Evaluation Report.

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