ML25175A076

03-11-80 Recommendations of the NRC Task Force on Bulletins and Orders

ML25175A076
Person / Time
Issue date:	03/11/1980
From:	Plesset M Advisory Committee on Reactor Safeguards
To:	Ahearne J NRC/Chairman
References
Download: ML25175A076 (1)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 March 11, 1980 Honorable John F. Ahearne Chairman

u. S. Nuclear Regulatory Co:rmnission Washington, OC 20555

SUBJECT:

REC~TIOOS OF THE NRC TASK FCRCE CN BULLETINS AND ORDERS During its 239th meeting, March 6-8, 1980, the Advisory Co:rmnittee on Reactor Safeguards completed a review of the recommendations of the NRC Task Force on Bulletins and Orders, hereafter called the Task Force.

'lhe ~

Subcom-mittee on 'lMI-2 Accident Bulletins and orders met with representatives of the NRC Staff and Utility CMners Groups on July 9, 1979, August 2, 1979, January 3-4, 1980, and March 4, 1980.

'lhe ACRS previously met with repre-sentati ves of the Task Force at the Committee's meetings of October 4-6, 1979, January 10-12, 1980 and February 7-9, 1980.

'lhe Task Force, fonned in May 1979, was charged with reviewirr;J and directing the 'lMI-2 related staff activities associated with the NRC I&E Bulletins, Commission Orders, and generic evaluations of loss of feedwater transients and small-break loss-of-coolant accidents for all operating plants to assure their continued safe operation. Specific review areas included systems reliability, vendor analysis methods and operating guidelines, plant procedures, and operator training.

'lhe results of the Task Force efforts have been reported in NUREX'i-0645, Volumes I and II, and a series of vendor specific reports noted below.

In its review, the Committee notes that the recommendations in reports NUREX'i-0565, 0611, 0623, 0626, and 0635 are those deemed by the Task Force to make the operatirr;J light W!lter reactor plants less susceptible to core damage during accidents and transients which are coupled with systems failures and operator errors.

'lhe Task Force has proposed that both the recommendations and the responsi-bility for their implementation be included in Section II.K.3 of NUREX'i-0660,

• NRc Action Plans Developed As a Result of the 'lMI-2 Accident*. 'lhe Commit-tee agrees with this course of action.

With regard to the recommendations the Committee has the followirr;J camnents:

• Reactor Coolant Pump Trip and High Pressure Injection (HPI)

Tennination Criteria: The NRC Staff has required prompt trip 1692

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Recent transients at some operating plants have resulted in RCP trip for mn-LOCA events and, in some cases, the use of the NRC awroved procedures for HPI termination have resulted in PORV or safety valve actuation due to overfilling of the primary system. 'lbe NRC Staff should, in conjt.mction with the licensees, review the criteria for HPI termination and reactor coolant pump trip to reduce unnecessary challenges to the pressurizer safety valves and prevent t.mnecessary trips of the reactor coolant pmtps which may increase the difficulty in establishing t.minterrupted core cooling.

• Feed-and-Bleed Cooling of the Primary System:

At the March 4, 1980 Subcanmittee meeting, the NRC Staff said that there are presently no requirements for the use of feed-and-bleed cooling for decay heat removal.

'lbe Comm! ttee believes that the availability of a diverse heat removal path such as feed and bleed is desirable, particularly if all secondary-side cooling is unavailable.

'lbe ACRS has established an Ad Hoc Subcanmittee to review this matter.

• Reduction of Challenges to the PORVs in B&W Plants:

As a result of the 'IMI-2 accident, the NRC Staff has required that all B&W plants raise the PORV actuation setpoint and l~r the high-pressure reactor trip setpoint in order to reduce the m1nber of challenges to the PORV.

While recent B&W operating reactor experience indicates that the PORV challenge rate has been reduced, there has been a corresponding increase in the nunber of reactor scrans. 'lbe Committee notes that an increase in the scram rate increases the probability of a deleterious impact on safety, and recommends that the NRC Staff continue to evaluate the overall impact of the above action on plant safety.

• Potential Unreviewed Safety Question with Regard to Automatic Initi-ation of the Auxiliary Feedwater System:

Several utilities have raised the issue of a potential unreviewed safety question with regard to automatic initiation of the AFW system, in the event of a main steamline break inside containment.

This issue should be reviewed.

'lbe Task Force has recanmended that the vendor methods used for snall break LOCA analysis should be revised, documented and sutxnitted for NRC review, and that plant specific calculations using NRC awroved methods should be provided thereafter.

'lbe NRC Action Plans also include an i tern which recanmends that the NRC develop and issue a position on required conservatisns in snall break calculations.

'lbe Committee believes that the schedule used for developing a revised NRC awroach to small break calculations should, if practical, be made canpatible with the schedule required of the NSSS vendors for revising their snall break models.

'Ibis 1693

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'Ibis implies some increased flexibility in the schedule.

With regard to the schedules proposed for the implementation of these recanmendations, the Committee believes that the orderly and effective implementation and the appropriate level of review and approval by the NRC staff will require a somewhat more flexible, and in some cases more extended, schedule than is implied by the Task Force reports.

'lbe Committee is still reviewing the NRC Action Plans 'Abich 'Nie lD'lderstand will include the Task Force's recanmendations discussed above, as '#811 as many other recommendations.

References:

Sincerely, Milton S. Plesset Olairman

1.

U.S. Nuclear Regulatory Commission, *Generic Evaluation of Small Break Loss-of-coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants", lENRC Report NUREx:i-0565, January 1980.

2.

U.S. Nuclear Regulatory Commission, *Generic Evaluation of Feedwater Transients and Small Break Loss-of-coolant Accidents in Westinghouse-Designed Operating Plants*, lENRC Report NUREx:;-0611, January 1980.

3.

U.S. Nuclear Regulatory Coounission, *Generic Assessment of Delayed Reactor Coolant Ptmtp Trip Dlring Small Break Loss-of-Coolant Accidents in Pressurized water Reactors", lENRC Report NUREx:i-0623, November 1979.

4.

U.S. Nuclear Regulatory Commission, "Generic Evaluation of Feedwater Transients and small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications", USNRC Report NUREx:i-0626, January 1980.

5.

U.S. Nuclear Regulatory Commission, "Generic Evaluation of Feedwater Transients and small Break Loss-of-coolant Accidents in Comrustion Engineering Designed Operating Plants", USNRC Report NUREG-0635, January 1980.

6.

U.S. Nuclear Regulatory Commission, "Report of the Bulletins and Orders Task Force", lENRC Report NUREx:;-0645, Volumes I-II, January 1980.

7.

U.S. Nuclear Regulatory Commission, "NRC Action Plans Developed As a Result of the TMI-2 Accident", USNRC Report NUREG-0660, Draft 3, March 5, 1980.

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