ML19350C868
| ML19350C868 | |
| Person / Time | |
|---|---|
| Issue date: | 01/11/1979 |
| From: | Fraley R Advisory Committee on Reactor Safeguards |
| To: | Bradford, Gilinsky V, Hendrie J NRC COMMISSION (OCM) |
| Shared Package | |
| ML19350C865 | List: |
| References | |
| ZECH, NUDOCS 8104060919 | |
| Download: ML19350C868 (4) | |
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6 Ef1 CLOSURE 4
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UNITED STATES 3
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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WASHINGTON. D. C. 20555 JAN 111979 MEMORANDUM FDR:
Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy Commissioner Bradford commissioner Ahearne FRQ4:
R. F. Fraley, Executive Director, ACRS
SUBJECT:
USE OF WASH-1400 BY TfE ADVISORY CQe1ITTEE ON REACTOR SAFEGUARDS
1he Office of the General Counsel has rotified this office that the Commissioners hne requested information regardire the use of the Reactor Safety Study, WASH-1400 by its advi-sory ccamittees, boards and panels.
Attached for your infonnation and use is a brief su:mrary of the manner in which the ACRS has been making use of WASH-1400 in its activites.
, Members of the Committee have contributed to and the ACRS Chairman has concurred in the attached.
R. F. Fraley Executive Director
Attachment:
- Applications.of WASH-1400 Methodology or Conclusions by ACRS dated 12/11/78 cc:
S. G ilk, SECY, w/att.
W. Shields,:OGC, w/att.
Contact:
R. Fraley, ACRS 4-3265-810406091{'
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12/11/78
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APPLICATIONS OF WASH-1400 METHODOLOGY OR CONCLUSIONS BY ACRS 1.
The Committee observed in its reports of April 8, 1975. July 14, 1976, and December 16,1976 (attached) that the methodology of WASH-1400 is useful for purposes of identifying important acci-dent sequences and for attempting to develop comparative and quan-titative risk assessments for low probability high-consequence accidents.
It noted, however, that the methodology cannot guar-antee that all major contributors to risk will be identified and a considerable element of judgment is required in assigning many of the-input. parameters. The Committee concluded that a substantial effort.would be required to develop and apply dependable methods for quantitatively accounting for the very large number of multiple correlate'd or dependent failure paths and to obtain the necessary failure rate data bases.
WASH-1400 did not cause the Corraittee to alter its judgment that
' reactors under construction or in operation do not represent an undue risk to the health and safety of the public nor did it result in any relaxation of ACRS conclusions or practices concern-ing Reactor Safety.
2.
WASH-1400 provided increased insight into containment failure modes following a postulated core melt and provided an improved basis for evaluation of the possibility of Class 9 accidents and the range G
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of consequences. This reinforced ACRS interest in the Generic Liquid Pathway Study which compared floating and land-based nu-clear facilities.
It led also to ACRS. interest in the possible development of the filtered vented containment as an improved safety system.
3.
More recently the Committee endorsed further development of the CRAC code for use in certain site evaluations.
CRAC is a compu-tation model developed for, and used in WASH-1400 to evaluate the consequences of serious accidents but does not directly involve the basic fault-tree / event-tree technique, nor the system relia-bility findings in WASH-1400.
4.
Individual Committee members have sometimes used WASH-1400 as a point of departure for questions, comments, or suggestions regard-ing safety related matters.
For exampic:
a.
It was suggested that the backfit decision-making process would be improved by using WASH-1400 methods to assess the reliability of alternate system designs.
b.
Preliminary coments on a recent staff ATWS study (NUREG-0460) were aimed at making direct comparisons between plant designs and the reliability goals in WASH-1400 (apparently one of the bases-of the report) more direct.
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The ACRS has used some results from WASH-1400 as a partial basis for requesting further evaluation of the adequacy of specific systems. For example, the ACRS has had a long-time interest in the capability of plants to survive safely a considerable loss of all AC power for an extended period. WASH-1400 indicated that the probability of a loss of all AC power was nonnegligible. The ACRS has asked the NRC staff for a comprehensive evaluaticn of the matter, including the possible need for design modifications. With the availability of WASH-1400 methodology and data, the ACRS was able to request an NRC Staff evaluation of the adequacy of the reliability of auxiliary feedwater and other systems of current detign.
6.
The consequence studies in WASH-1400 provided additional background
'information for ACRS consideration of emergency preparations.
Attachments:
'l.
Ltr. to W. A. Anders dtd. 4/8/75 2.
Ltr. to M. K. Udall dtd. 7/14/76 3.
Ltr. to M. K. Udall dtd. 12/16/76 t
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WASHINGTON. D. C. 20555 July 14,1976 n e Bonorable Morris K. Udall, ch ircan Subecmuittee on Energy and the Environ::ent Comittee on Interior and Insular Affairs United States House of Representatives F,ashington, DC 20515
Dear Congress:
nan 03:11:
At its 195th meeting on July 8-10, 1976, the ASvisory rrmnittee on Iteactor Safeguards (ACPS) considered the points raised in pur June 14, 1976, letter on the Reactor Safety Study (RSS, MSH-1400, WPIG 75/014). Tne ACPS reviewed the draft version of the Reactor Safety StuSy in late 1974 an5 c-,rly 1975 and suhaitted a report to the Nuclear Regulatory Whh on 2prd1 B,1975.
A copy of the ACRS reprt is attached.
Your letter identified ' eleven issues on which you recpestea w.ut and the Conraittee is pleased to respond to itsces 1, 3, 4, 5, B, 9 ac510...However, extensive tire and effort muld be required try the ACES to cesponS adequately to the other topics and the needed effort vould have to 1:e fenred.into overall considerations of other ACPS functions, incln95ng onedletary review of applications for construction pnmits and operating 'Hmn=e for comrercial nuclear pcuer plants.
'lhe Caznittee's respnses follow:
1.
"lhe extent that the NURE 75/014 fault-tree analysis mm to under-
' standing of the likelihoca of major nuclear rme t -Monts."
'lhe ACPS believes that the fault-tree methodology nwl in t'ha 3teactor Safety Study to develop comparative and quantitative risk -me for postulated accident sequences represents a valuable contribaHnn to the wma=rstanding of the likelihood of major nuclear reactor n e-ia nts.
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- 3. " Adequacy of data base for EUREG 75/014 type fault-tree anzilysin."
As noted in our reprt of April 8,1975, the ACPS bWues tthat a 'better data base will be required to evaluate the validity of the MSS'.s xpantitative estimate _s of the likelihood of low probability high -muence events, and recoc:cends that current efforts to compile, categorize and evaluate nuclear and other applicable industrial experience Le c#mAM in !brnaSth :and depth to improve the data base for further studies of t+is tEpe.
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[ine Honorable i4 orris'K. Cdall Jtily 14,1976 4.
" Sensitivity of WRH3 75/014 conclusions to differences in reactor design, in site characteristics, in local ineteorological conditions and in population distributions."
All of the factors noted abave will have some ofrect on the probability or consequences of a serious accident. We Cormtittee has reco.nnended that the methodology of the Study be applied to other types and designs of reactors, other site conditions and other accident initiators and sequences.
If this is done, it will provide greater insight into the sensitivity of. differing reactor designs and safety features.
6.
" Adequacy of NUPS3 75/014 methodology to take account of gradual degradation of plant safety over plant lifetime."
he Comittee believes the nethodology is capable of taking into account wear out of components and degradation of equipnent over the lifetime of the plant but an appropriate data base needs to be developed.
8.
"Need for periodic updating of HUREG 75/014 to take account of new data."
We Comraittee believes that a continuing effort is desirable in the application of the methodology developed by the Reactor Safety Study not only to factor in new data but also to consider design variations and new concepts.
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9.
"Keed for continuing analysis of MURr 75/014 for ' purposes of delineating areas of research and data collection."
The Comraittee blieves that the EUREG 75/014 nethodology should be used to aid in delin' areas for further research. Special emphasis should be given t'o qum ation of the initiators, probabilities, and consequences of core rcelting.
- 10. "The extent to which NURCG 75/014 can be used to aid develo,m.ent of regulatory policies concerning design, construction, and operations."
he Committee has recommended to the NRC that many of the techniques used in the Study can and should be used by the reactor designers to improve scfety and by the NRC Staff as a supplement to their safety assessnent.
Sincerely yours, 5
Dade W. !beller Chairman I
Attadinent:
Ltr. to Hon. W. Anders frora D. W.
Foeller, dtd 4/8/75 re: msn-1400 6
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NUCLEAR REGULATORY COMMISSlQA ADI 3RY COMMITTEE ON REACTOR SAFEGL
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.{DS WASHINGTO?d, D. C. 20555 December 16, 1976 The Honorable Morris K. Udall, Chairman Subcomittee on Energy and the Environm2nt Comittee on Interior and Insular Affairs United States House of Representatives Washington, IX' 20515
Dear Congresaran Udall:
At itis 200th meeting, December 9-11, 1976, the Advisory Committee on Re-actor Safeguards (ACRS) continued its consideration of the points raised in your June 14, 1976, letter on the Reactor Safety Study (RSS, UhSH-1400, NUREG 75/014). The ACRS had previously considered these matters at its 196th and 199th meetings and had re ponded to issues 1,3,4,6,8,9 and 10 in its letter to you dated July 14, 1976. In its further con-sideration of the remaining four issues, the Comittee had the benefit of meetings of its Reactor Safety Study Working Group with the Nuclear Reg-ulatory Comission Staff in Washington, DC, on October 12, 1976, and I;
November 10, 1976.
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The ACRS is continuing to evaluate the considerable body of inf.Tation presented in the RSS report, its acpendices, and the corrents received on it, giving primary attention to the potential implications of the report for the reactor licensing proces's. This letter provides the Comittee on Interior and Insular Affairs a brief restrr.e of current ACRS thought on issues 2, 5, 7 and 11.
"2. Adequacy and appropriateness of analysis used in NUREG 75/014 for purposes of estimating the ]Delihood of low probability, high con-sequence events."
The ACRS believes that. the methodology of NUREG 75/014 is useful for purposes of identifying irrpodant accident sequences and for attempting to develop coriparative and quantitative risk assessments for low prob-ability, high-consequence accidents. However, the ACRS believes that considerable effort by more than a single group over an extended period of time will be required to evaluate the validity of the results in IUREG 75/014.in absolute terms. Among the matters tinich will warrant emphasis in such an evaluation are the following:
inproved quantification of acci-dent initiators; the identification and evaluation of atypical reactors; the influence of design errors; improved quantification of the role of operator errors; improved quantification of consequence modeling; and the.
- development of improved data for systems, co.Tponents and instrtrnents under.
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2-December 16, 1976 The Honorable Morris K. Udall The ACRS believes that NUREG 75/014 represents a very considerable con-tribution to the understanding of reactor safety and provides a point of departure for quantitative assesment.
"5.
Adequacy of NUREG 75/014 methodology to take account of multiple, correlated errors in procedures, design, judgment, and construction such as those leading to the Browns Ferry fire."
The ACRS believes that the rethodology of NUREG 75/014 is useful in ac-counting for that portion of the risk resulting from identifiable potential cort:non mode or dependent failures, and can be used to search out the pos-sibility of multiple correlated errors. However, the rethodology cannot guarantee that all cajor contributors to risk will be identified, and a considerable element of subjective judg:ent is involved in assigning rany of the quantitative input para :eters. Both for nuclear and non-nuclear applications, for cosplex systens, where multiple, correlated failures or co=on catise failures ray be significant, the record shoss that investi-gators wrking independently vill frequently make estimates of system unreliability which differ from one another by a large factor. At this
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stage of its review, the ACRS believes that a substantial effort may be required to develop and apply dependable methods for quantitatively ac-counting for the very large ntnber of multiple correlated or dependent failure paths and to obtain the necessary failure rate data bases.
Uhether multiple, correlated errors win dominate the overall risk, hea-ever, is subject to question, particularly if simpler postulated accident sequences are generally the dominant aantributors to the likelihood of system failure.
- 77. Extent to which the final version of NUREG 75/014 takes into account coments on the draft version.".
The ACRS is in the process of reviewing the disposition of selected coments received by the Reactor Safety Study Group, particularly as they have impli-
, cations for short or long-term improvements in reactor safety. The ACRS plans to continue this type of activity; however, it is beyond the scope
. or available working time of the ACRS to review in detail the extent to which the final version of NUREG 75/014 takes into account the comments received.
"11.
Validity of NUREG 75/014 conclusions regarding accident aansequences."
As stated in its report to you of July 14, 1976 and as indicated in its
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response to other questions in this group, the ACRS believes that consi-derably more effort on the part of various contributors is needed to p
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-3 A.ccher 16, 1976 evaluate the quantitative validity of NUREG 75/014 conclusions regarding accident consequences. Based on information currently available,. the ACRS would assign a greater uncertainty to the results than that given in IMREG 75/014.
The ACRS believes that the past and current practice of trying both to make accidents very improbable and to provide maans to cope with or ameliorate the effects of accidents has been the correct approach to nuclear reactor safety.
The ACES review of the Reactor Safety Stu3y has not caused the ACRS to alter its judgrent that operation of reactors now under construction or in oper-ation does not represent an undue risk to the health and safety of the public.
The ACRS believes that UUTGG 75/014 has suggested many fruitful areas for study and evaluation for potential ir: prover.ents in light uater power reactor safety. The ACRS also telieves that the extension of such risk assesnr.ent methoSology to the total spectrum of activities involved in the production of nuclear power and in the production of electric power by other means, as well as to' other technological aspects of society, could add significantly to our overall understanding of risk.
Sincerely yours, 9Og 3
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ADVISORT COMMITTEE ON REACTOR SAFhvUARDS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 April 8, 1975
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Honorable Uillian A. Anders Chaircan U. S. Ilucicar Regulatory Connission Washington, D. C.
20555
Subject:
REACTOR SAFETY STUDY, UASH-1400 Since the release of the draft Reactor Safety Study, UASH-1400 (RSS) in August 1974, the Advisory Connittee on Reactor Safeguards
, has been revictrin3 the considerabic body of inforcation presented in the report, its appendices, and the connents received on it, giving prinary attention to the potential inplications of the draf t report on the reactor licensing process.
In its revicu, the Connittee has had the benefit of Subconnittee ricecings held on October 9, Novenbar 22, and Decenber 20, 1974, and March 5, 1975, and of full Co=nittee nectings held on October 10-12, October 31-Novenber 2, November 14-16, Decenber 5-7, 1974, and January 9-11, February 6-8, March 6-8, April 3-5,1975.
The ACRS believes that the RSS represents a valuable contribution to the understanding of light unter reactor safety in its categorization of hypothetical accidents, identification of potential ueak links for the tuo reactors studied, and its efforts to develop conparative and quantitative rish assessnents for accident sequences exanined The Connittee helieves that a continuing effort and better data vill be required to evaluate the validity of the quantitative results in absolute terns. Special enphasis should be given to quantification of the initiators, probabilitics, ant' consequences of core celting.
.The Coccittee believes that the methodology of the RSS should be applied to other types and designs of reactors, other site conditions and other accident initiators and sequences, and that the currect
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efforts to compile, categorize, and evaluate nuclear experience should be extended in breadth and depth to inprove the data base for future studies of this type.
The Connittee believes, further, that the RSS can serve as a model for sinilar studies of the failure probabilities, consequences, and resulting risks of other hazards (both nuclear and non-nucicar) to-the health and safety of the public.
The Connittee believes that nany of the techniques used in the
, RSS can and should be used by reactor designers to improve safety and by the NRC Staff as a supplement to safety assessoent.
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Honorable WJ111am A. Anders April 8, 1975 The Cotzsittee's review of the RSS has not caused the Comnittee to alter its judgecent that reactors now under construction or in operation,do not represent undue risks to the health and safety of the public.
The Cocaittee vill continue 'to review the RSS and vill co=nent further on it in the future.
Sincerely, Original Sissalby;
. hV. E m
%d William Kerr
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