ML19309B252
| ML19309B252 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/11/1980 |
| From: | Plesset M Advisory Committee on Reactor Safeguards |
| To: | Ahearne J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19309B249 | List: |
| References | |
| NUDOCS 8004030333 | |
| Download: ML19309B252 (3) | |
Text
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'c; UNITED STATES NUCLEAR REGULATORY COMMISSION E
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 2.I k
/
WASHINGTON, D. C. 20555 March 11, 1980 Honorable John F. Ahearne Chairman U. S. Nuclear Regulatory Commission Wr.shirgton, DC 20555
Dear Dr. Ahearne:
SUBJECT:
RECCEMENDATIONS OF THE NRC TASK FCRCE CN BULLETINS AND ORDERS During its 239th meeting, March 6-8, 1980, the Advisory Committee on Reactor Safeguards completed a review of the recommendations of the NRC Task Ibrce on Bulletins and orders, hereafter called the Task Force.
We ACRS Subcom-mittee on TMI-2 Accident Bulletins and Orders met with representatives of 3,
i the NRC Staff and Utility Ovners Groups on July 9, 1979, August 2,1979, January 3-4, 1980, and March 4,1980.
We ERS previously met with repre-sentatives of the Task Force at the Committee's meetings of October 4-6, 1979, January 10-12, 1980 and February 7-9, 1980.
We Task Force, formed in May 1979, was charged with reviewing and directing the TMI-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.
We results of the Task Force efforts have been reported in NURD3-0645, Volumes I and II, and a series of vendor specific reports noted below.
In its review, the Committee notes that the recommendations in reports NUREG-0565, 0611, 0623, 0626, and 0635 are those deemed by the Task Force to make the operating light ster reactor plants less susceptible to core damage during accidents and transients which are coupled with systems failures and operator errors.
We Task Force has proposed that both the recommendations and the responsi-bility for their implementation be included in Section II.K.3 of NUREG-0660, "NRC Action Plans Developed As a Result of the TMI-2 Accident". W e Commit-tee agrees with this course of action.
With regard ~ to the recommendations the Committee has the following comments:
l
' Reactor Coolant Pump Trip and High Pressure Injection (HPI)
Termination Criteria: We NRC Staff has required prcrnpt trip 8004030 %
4 Honorable John F. Ahearne March 11, 1980 of the reactor coolant pu.nps in the event of a sull-break I4CA.
Recen, transients at some operating plants have resultt3 in RCP trip for non-LOCA events and, in some cases, the use of the NRC approved procedures for HPI termination have resulted in PORV or safety valve actuation due to overfilling of the primary system. We NRC Staff should, in conjunction 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 unnecessary trips of the reactor coolant punps which may increase the difficulty in establishing uninterrupted core cooling.
' Feed-and-Bleed Cooling of the Primary System:
At the March 4,1980 Subccrnmittee meeting, the NRC Staff said that there are presently no requirements for the use of feed-and-bleed coolirg for decay heat removal.
We Committe wlieves that the availability of a diverse t
heat removal cath such as feed and bleed is desirable, particularly if s
all secondar, -side cooling is unavailable.
We ACRS has established an Ad Hoc Subconmittee to review this matter.
'Redtx: tion of 01allenges 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 lower the high-pressure reactor trip setpint in order to reduce the number 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 ntsnber of reactor scrans. We 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.
We Task Force has recommended that the vendor methods used for small break IDCA analysis should be revised, documented and subnitted for NRC review, and that plant specific calculations using NRC approved methods should be provided thereafter.
%e NRC Action Plans also include an iten
~
which recommends that the NRC develop and issue a position on required conservatisms in small break calculations.
%e committee believes that the schedule used for developing a revised NRC approach to small break calculations should, if practical, be made compatible with the schedule required of the NSSS vendors for revising their snall break models.
his
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Honorable John F. Ahearne Parch 11,1980 should lead to a more efficient use of available resources and rray let.d to an earlier developnent of improved analyses.
his implies some increased flexibility in the schedule.
With regard to the schedules proposed for the implementation of these reccmnendations, the Committee believes that the orderly and effective implementation and the appropriate lwel of review and approval by the NRC staff will require a some hat more flexible, and in some cases more I
extended, schedule than is implied by the Task Force reports.
We Committee is still reviewing the NRC Action Plans which we understand will include the Task Force's reccxnmendations discussed above, as wall as many other recommendations.
Sincerely, Milton S. Plesset Chairman
References:
t 1.
U.S. Nuclear Regulatory Commission, " Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants", USNRC Report NUREI3-0565, 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", WNRC Report NUR m-0611, January 1980.
3.
U.S.
Nuclear Regulatory Commission, " Generic Assessment of Delayed Reactor Coolant Pump Trip During Small Break Loss-of-Coolant Accidents in Pressurized Water Reactors", USNRC Report NURD3-0623, 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 NURS3-0626, January 1980.
5.
U.S.
Nuclear Regulatory Commission, " Generic Evaluation of Feedwater Transients and Small Break Ioss-of-Coolant Accidents in Combustion Engineering Designed Operating Plants", USNRC Report NUREG-0635, January 1980.
6.
U.S. Nuclear Regulatory Commission, " Report of the Bulletins and Orders Task Force", USNRC Report NUR E-0645, Voltrnes I-II, January 1980.
l.
7.
U.S. Nuclear Regulatory Commission, "NRC Action Plans Developed As a l
Result of the 'D4I-2 Accident", USNRC Report NUREG-0660, Draft 3, i
March 5, 1980.
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