ML19305D546
| ML19305D546 | |
| Person / Time | |
|---|---|
| Issue date: | 03/11/1980 |
| From: | Plesset M Advisory Committee on Reactor Safeguards |
| To: | Ahearne J NRC COMMISSION (OCM) |
| References | |
| ACRS-R-0864, ACRS-R-864, NUDOCS 8004150208 | |
| Download: ML19305D546 (3) | |
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UNITED $9AM33 N5b dbY I
NUCLEAR REGULATORY COMMISSION y
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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s CASHINGToN, D. C. 20655
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a, March 11, 1980 Honorable John F. Ahearne j
Oiairman j
U. S. Nuclear Regulatory Comission Washington, DC 20555
Dear Dr. Ahearne:
SUBJECT:
RE'XMMENIRTI(NS OF ME EC TASK FCRCE CN BULLETINS AND ORDERS During its 235th meeting, March 6-8, 1980, the Advisory Committee on Reactor -
Safegmrds competed a review of the recommendations of the EC Task Pbrce on Bulletins ano Orders, hereafter called the Task Pbrce. % e ACRS Subcom-mittee on MI-2 Accident Bulletins and Orders met with representatives of the EC Staff and Utility Owners Groups on July 9,1979, August 2,1979, l
January 3-4, 1980, and March 4,1980.
%e ACRS previously met with repre-sentatives of the Task Force at the Comittee's meetings of October 4-6, 1979, January 10-12, 1980 and February 7-9, 1980.
he Task Force, formed in May 1979, was charged with reviewing and directing i
the MI-2 related staff activities associated with the NRC IEE Bulletins, i
cournission Orders, and generic evaluations of loss of feedwater transients l
and small-break loss-of-coolant accidents for all operating plants to assure their continued safe operation.
Specific review areas included l
systems reliability, vendor analysis methods and operating guidelines, l
plant procedures, and operator training.
We results of the Task Pbree efforts have been reported in NUREG-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 NUREU-0565, 0611, 0623, 0626, and 0635 are those deemed by the Task Force to make the operating light water reactor plants less susceptible to core damage during accidents and transients which are coupled with systems failures and operator errors.
%e 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, j
"NRC Action Plans Developed As a Result of the MI-2 Accident". We Commit-tee agrees with this course of action.
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l With regard to the recommendations the Committee has the following comments:
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' Reactor Coolant Pump Trip and High Pressure Injection (HPI)
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Termination Criteria: We NRC Staff has required prompt trip l
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p.
_ _. _ _..__ _ _... _ _ _ __ __ _ __ __ _ __ _.8 0.0A 15.0 2.0 8
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Honorable John F. Ahearne March 11, 1980 of the reactor coolant pisnps in the event of a maall-break IDCA.
Recent transients at some operating plants have resulted in RCP trip for non-LOCA events and, in some cases, the use of the EC approved
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procedures for HPI termination have resulted in PORV or safety valve l
actuation due to overfilling of the primary system. %e EC Staff should, in conjunction with the licensees, review the criteria for HPI I
termination and reactor coolant pump trip to reduce unnecessary challenges to the pressurizer safety valves and prevent mnecessary trips of the reactor coolant punps Wich may increase the difficulty l
in establishing minterrupted core cooling.
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' Feed-and-Bleed Cooling of the Primary System:
At the March 4,1980 i
Subecznmittee meeting, the EC Staff said that there are presently no requirements for the use of feed-and-bleed cooling for decay heat renoval.
We committee believes that the availability of a diverse l
heat renoval path such as feed and bleed is desirable, particularly if l
all secondary-side cooling is unavailable.
he ACRS has established an Ad Hoc Subcommittee to review this matter.
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' Reduction of Challenges to the PORVs in B&W Plants:
As a result of the 'IMI-2 accident, the EC Staff has required that all B&W plants raise the PORV actuation setpoint and lower the high-pressure reactor trip setpoint in order to reduce the number of challenges to the PORV.
l While recent B&W operating reactor experience indicates that the PORV t
challenge rate has been reduced, there has been a corresponding l
increase in the number of reactor scrams. Se Oxumittee notes that an l
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
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raised the issue of a potential unreviewed safety question with
. regard to automatic initiation of the AEW system, in the event of a main steamline break inside containment.
%is issue should be reviewed.
I he Task Force has recommended that the vendor methods used for small I
break IDCA analysis should be revised, documented and submitted for NRC review, and that plant specific calculations using NRC approved methods should be provided thereafter.
We NRC Action Plans also include an iten sich recommends that the NRC develop and issue a position on required conservatisms in anall break calculations.
We Committee believes that the schedule used for developing a revised EC approach to small break calculations should, if practical, be made compatible with the schedule required of the NSSS vendors for revising their anall break models.
his e
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c Honorable John F. Ahearne Forch 11, 1980 i
should lead to a more efficient use of available resources and may lead to
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an earlier developnent of improved analyses.
his implies some increased j
flexibility in the schedule.
With regard to the schedules proposed for the implanentation of these i
recommendations, 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 /
l extended, schedule than is implied by the Task Pbrce reports.
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he Committee is still reviewing the EC Action Plans which we understand j
will include the Task Force's recomendations discussed above, as well as j
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many other recommendations.
l l
Sincerely, Milton S. Plesset
(
Olairman 1
References:
1.
U.S. Nuclear Regulatory Comission, " Generic Evaluation of Snall Break
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Mss-of-coolant Accident Behavior in Babcock & Wilcox Designed 177-FA l
Operating Plants", USNRC Report NIREG-0565, January 1980.
2.
U.S. Nuclear Regulatory Commission, " Generic Evaluation of Feedwater Transients and Small Break kss-of-Coolant Accidents in Westinghouse-Designed Operating Plants", USNRC Report NLREG-0611, January 1980.
3.
U.S.
Nuclear Regulatory Comission, " Generic Assessment of Delayed Reactor Coolant Pump Trip During Snall Break kss-of-Coolant Accidents in Pressurized Water Reactors", USNRC Report NUREG-0623, Novenber 1979.
t 4.
U.S. Nuclear Regulatory Comission, " Generic Evaluation of Feedwater Transients and Snall Break kss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications", USNRC l
Report NURB3-0626, January 1980.
5.
U.S. Nuclear Regulatory Commission, " Generic Evaluation of Peedwater Transients and Small Break kss-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 NURD3-0645, Volunes I-II, January 1980.
7.
U.S. Nuclear Regulatory Comission, "NRC Action Plans Developed As a
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Result of the 'IMI-2 Accident", USNRC Report NUREG-0660, Draft 3, March 5, 1980.
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