ML20235A875

Summarizes ACRS 336th Meeting on 880407-09 Re Efforts to Devise Methods for Implementing Safety Goal Policy Enunciated by Commission in 1986

ML20235A875
Person / Time
Issue date:	04/12/1988
From:	Kerr W Advisory Committee on Reactor Safeguards
To:	Zech L NRC COMMISSION (OCM)
Shared Package
ML20235A877	List: ML20235A875 ML20235A879 ML20235A988 ML20235V448 ML20235V614
References
ACRS-R-1341, NUDOCS 8804210387
Download: ML20235A875 (5)

Text

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• • • • • April 12, 1988 i

The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, DC 205S5

SUBJECT:

PROGRAM TO IMPLEMENT THE SAFETY GOAL POLICY -- ACRS COMM During the 336th meeting of the Advisory Comittee on Reactor Safe-guards, April 7-9, 1988, and in previous meetings, we discussed NRC Staff efforts to devise methods'for implementing the Safety Goal Policy enunciated by the Comission in 1986. Following our report to you of May 13,1987, "ACRS Comments on an Implementation Plan for the Safety Goal Policy," the Comission directed the Staff to develop a plan along the lines we had suggested.

The ACRS and its Subcommittee on Safety Philosophy, Technology, and Criteria have met several times with members of the Office of Nuclear Regulatory Research (RES) to discuss means by which the implementation plan suggested in our May 1987 report might be " fleshed out." 'We expect that additional meetings will be necessary over the next few months.

Our understanding is that the Staff will then document a description of its program. We believe it would be useful to provide you with some interim comments. These follow:

Definition of a "Large Release" The Safety Goal Policy includes a general performance guideline that there should be a probability no greater than IE-6 per reactor-year Exactly whatof is a

large release from any operating nuclear power plant.

meant by a large release was not defined in the Policy, but it has been suggested that a definition is needed in the implementation plan.

We believe the definition of a large release should make clear that a release to the biosphere of a substantial fraction of the core inventory is intended. In particular, it is misleading to define it in tems of Of health effects; those goals are stated elsewhere in the Policy.

course, in assessing whether the 1E-6 goal is met by the body of regu-latory rules and practice, one will have All to consider and evaluate of the probabilistic risk-specific sequences for individual plants. assessments (FRAs) pr versally agreed that the " bottom line" estimates Nonetheless, thereby derivedof a definition area among the weakest results of a PRA.

Attachment 3

The Honorable lendo W. Zech, Jr. April 12, 1988 1arge release in tems of a number of curies released would provide a durable objective for the calculations, which will improve with time, and which must be done as best one can at any given time. In the end, of course, a particular number will have to be specified, and it must be consistent with the other elements of the program.

Examples of what we have in mind might be helpful. We regard the release to the environmmt that occurred at THI-2 in 1979 as not a large release. The release that occurred in the accident at Chernobyl Unit 4 in 1956 was a large release. The fact that there were apparently no

" prompt" radiological fatalities among the offsite population at Chernobyl is irrelevant under the proposed definition.

Definition of " Core Melt" In our report of Pay 13, 1987, we suggested a perfomance objective for

" prevention" systems as a calculated core melt probability of less than IE-4 per reactrr-year. Exactly what is meant by core melt, seemingly a s

simple cuestien, presents a prcblem for analysts and others considering the details of ruclear power plant accidents. The most likely sequence of core overheating, melting, and displacement can be viewed as a hypothetical sequence of events. Each e' vent is less likely than that preceding it because the sequence may be interrupted at any point, for example, by succ?ssful perfomance of emergency procedures. These events might be defined as:

(1) loss cf adequate core cooling (core overheating beyond design-basis limits),

(2) onset of significant damage to the core, (3) and displacement of the core within the reactor pressure melting vessel (as in the THI-2 event),

y (4) passage of molten core out of the reactor pressure vessel (e.g.,

" core on the floor *).

We would apply the 3E-4 objective to the first event in this sequence, with the expectation that there is a significant but undefined margin in likelihood between it and the remaining events. We note that referring to this as a " Core Melt Perfomance Objective" is an unfortunate choice; however, it is one that is well established in the nuclear safety community. 4

i The Honorable Lando W. Zech, Jr. April 12, 1988 l

Definition of the Plant Performance Objective Our May 1987 report reconnended that a performance objective expressing "how well the plant is operated" should be developed. The RES Staff has indicated that it doesn't know how to do this and plans no further work regarding this matter. Without a perfomance objective of this sort. l the severe gap in the logic of the Safety Goal Policy, which led to our original recommendation, remains.

The problem is as follows:

. The Safety Goal Policy is intended, as we understand it, to be a declaration of intent about how safe operating nuclear plants are to be. However, PRA, the primary tool by which performance of .

- plants against the safety goal is to be judged, uses few data on operational performance. Most of the analyses in a PRA depend on attributes of the plant design. Very little information about how e plant is actually operated is used. Where actual operational performance is included (e.g., equipment failure rates and pre-dictions of cperator response), most of the data used are generic to industry experien;e and little reflect attributes of the oper-ation of the particular plant being . analyzed. This is really an inherent weakness in the present art of PRA.

. Although no means are presently apparent for incorporating a more complete definition of operational performance into the Policy implementation, we believe credibility of the Policy suffers without it. Decu rch cnuld belo. It might be possible to somehow better incorporate attributes of operational performance into PRA.

If this cannot be done, a prominent caveat, e.g., a warning that PRA results do not tell the full story, should be made a part of the Policy or of the implementation plan.

We note that at one of our recent meetings, Nuclear Reactor Regulation Staff described piens for further work having as its objective a better description (for use in PRAs, e.g.) of the contribution to risk or to safety made by the plant operating staff. Results of this work could contribute to the formulation of a performance objective.

Use of Cost-Benefit Analysis Cost-benefit analysis has a role in regulatory practice under the backfit rule. In this context, the role of the safety goal should be only to help provide a definition of what is meant by ' adequate for safety." If it is found that a regulation is permitting plants to be licensed which seem to have safety perfomance poorer than the guidance provided by the safety goal, then that regulation should be revised, without recourse to cost-benefit arguments.

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The Honorable Lando W. Zech, Jr. April 12, 1988 Need For Review of Regulations From the Perspective of the Safety Goal Policy There is a need to consider wnat is meant by implementation of the Safety Goal Policy. We have suggested that the Policy not bt "used to make narrowly differentiated decisions about specific plants." Instead, it should be used as a primary means for judging the suitability and necessity of specific regulations and regulatory practices. We include practices in this discussion because many of the requirements levied by 7 the NRC on licensees are not part of formal regulations but proceed from a acre informal bedy of practice. We put aside, for now, the question of whether this in itself is or is not a problem but suggest only that the informal, as well as the formal, regulatory practices should be constrained by the Safety Goal Policy.

The next question is whether the Policy should be used only reactive 1) in assessing proposed regulatory changes evolving from other programs (e.g., new requirements coming from resolution of a USI), or should be used more actively in assessing the present body of regulatory practice.

We recommend the latter.

The existing body of regulations and regulatory practice has grown enormously over the past 30 years. This growth has been largely a bottom-up process as the regulatory staff and ACRS have reacted to proposals from applicants and vendors and responded to developing technical information and plant experience. We believe it is possible to make a zero-based assessment of this body of regulations to deter-mine: (a) which parts are contributing effectively to assure that plants are appropriately safe, (b) which parts are unnecessary, and (c) which parts need to be strengthened or better focused.

It is the responsibility of the NRC to move in the direction of such an assessment. It will not be easy but should begin now for several reasons:

There is a hiatus in applications for new plants.

There is now an extensive body of experience with operation and regulation that did not exist 30 years ago.

There is now much more complete information about severe accidents than existed previously.

. PRA has matured and is available as a tool.

And finally, the Safety Goal Policy is available as a thoughtful and agreed-upon measuring criterion.

4 The Honorable Lando W. Zech, Jr. April 12,:1988 l

We trust the above coments will be useful as the Staff continues with development of the Pclicy implementation.

Additional remarks by ACRS Member Harold W. Lewis are presented below.

Sincerely, W. Kerr Chairman Additional Remerks By ACRS Member Harold W. Lewis The Committee has, in its May 1987 report, defined the term " core melt"  ;

to mean loss of assured core cooling which can result in severe core  ;

damage, to me,tch the probability objective of IE-4 per reactor year. In this report the definition is made even more restrictive. While there is ambiguity in the community about the meaning of the term (as noted by  ;

the Committee), the redefinition has an enomous impact on the effect of  :

the goal. There is a considerable difference of probability between loss of adequate core cooling and melting of the core, the former more probable but not necessarily damaging. Since the assignment of the term

" core melt" to an event which need not melt the core unnecessarily biases the interpretation of the safety goal. I believe it is the job of  ;

the Commission to clarify what is meant by the tem, rather than for the  !

Comeittee to read minds. I take the simplistic view that a core melt requires a rolten core. In the law, the established procedure for resolving apparent ambiguities is to start with the plain meaning of the words.

I also believe it important that the Comission (not the Staff) clarify its intent in promulgating the Safety Goal Policy. Though the goals were stated by the Comission two years ago, we continue to hear of Staff actions which are justified by one or another version of "if we can see a way to improve safety, we will." Presumably the Comission, by giving an answer to the how-safe-is-safe-enough question, intended '

precisely to dampen such unbounded enthusiasm. I believe the Comission should reinforce its guidance to its staff.

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,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20856 44 e..,*

f May 10, 1988 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. fiuclear Regulatory Conmission Washingten, D.C. 20555

SUBJECT:

FIRE RISK SCOPING STUDY

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Ir. our July 16, 1986 letter to the Commission concerning fire protection issues, we urried reconsideration of the budget and manpower allocations tc the fire-related portions of the NRC safety research program. In response, in 6 memorandun dated July 24, 1986, Chairman Zech recommended that the Staff work closely with the ACRS to assess further research needs and to consider what priority should be given to fire protection rese6rch. The Staff acted in January 1987 by initiating the Fire RirA Scoping Study at the Sandia National Laboratories (SNL), and we provided our views on the scope and direction of this Study in a report to the Connissien dated August 10, 1987.

During our 337th meeting, May 5-7, 1988, we met with representatives from the Office of Nuclear Regulatory Research and SNL to discuss the results and conclusions of the Fire Risk Scoping Study. This matter was considered by our Subcommittee on Auxiliary Systems during a meeting on f4 arch 9, 1988. We also had the benefit of the document referenced.

We were informed that the Staff is now considering what actions should be taken regarcing the disposition of the recommendations resulting from the Study, and a decision is expected by the end of FY 1988. If some of the asserted results survive deeper scrutiny, they could be important.

Therefore, we recomend that the Staff evaluate the results and conclu-sions of the Study and decide on a course of action on a schedule which permits any high-priority research to be initiated in FY 1989. We wish to be kept inforrned of further developments, and we expect to provide coments after the Staff has identified its proposed plans.

Sincerely, W. Kerr Chainnan I 9 ? N Z.b?O N- Attachment 4

Mr. Victor Stello, Jr.

Reference:

Draf t Report dated March 1988, Sandia National Laboratories, NURE6/CR-5088, sat;D 88-0177, " Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, including Previously Unaddressed Issues"

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., * " * * / - Hay 10, 1980 The Honorable Lando W. Zech, Jr.

Ctairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Chairman Zech:

SUSJECT: PROPOSED GENERIC LETTEP GN INDIVIDUAL PLANT EXAMINAT THE PROPOSED INTEGRATED SAFETY ASSESSMENT PROGRAM II Durig the 237th meeting of the Advisory Comittee on Reactor Safe-guards, r;cy '-7,1988, we discussed a draft generic letter prepared by the NRC Staff as guidance for Individual Plant Examinations (IPEs) for We also discussed the proposed severe accident vulnerabilities.

Integrated Safety Assessment Program II (ISAP !!) and related inforna-tien. Both of these topics have been considered during previous meetings of the ACRS, and we reported our preliminary views on the IPE generic letter in our report of June 9,1987 and on the ISAP process in our report of July 15, 1987. The ACRS Subcommittee on Severe Accidents net on April 26, 1988 to discuss the latest version of the The ACRS Subcorsittee on Generic proposed generic letter on IPIs.to discuss ISAP II. We also had the Items met on April 27, 1988 benefit of discussions on both topics with members of the NRC Staff and industry representatives, as appropriate, and the availability of the documents referenced.

These two programs developed by different NRC Staff groups have not been integrated, even though they deal with many of the same issues.

is for this reason that we are providing our coments on both programs in a single letter. The present Staff positions, as we It understand them, are that the IPE generic letter should be issued in its present form and that implementation of the ISAP !! should not be We disagree with both of these positions.

pursued at this time.

Instead, we believe that the IPE program should be expanded to incor-porate all outstanding safety issues', not .just those under the severe accident rubric. The generic letter should be revised accordingly. '

The ISf.P II approach should then serve as the instrument by which changes in plant equipment or procedures identifi the NRC Staff.

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The Honorable Lando W. Zech, Jr. ,2- May 10,1988 -

We consider the most recent draft of the IPE generic letter an in-provement over that which we comented on in our report of June 9, 1987. However, in our report of March 15, 1988, we expressed our concern that there was a lack of coherence among the several principal regulatory programs of the Comission. We believe the IPE program offers an opportunity for providing improved coherence. In its present form, the generic letter will, instead, continue the current compartmentalization.

We believe that IPE and ISAP 11 can be recast in a reasonable time and with reasonable expenditure of resources. Radical changes are not recessary, but some modifications and improvements in focus are. We propose a program characterized as follows:

• The purpose of IPEs would be acknowledged as broader than the original intent of " searching for outliers." Instead, it would call for a general risk reassessment of each plant using the body of information available from the TMI-2 accident experience, development of PRA, existing severe accident research, and the general experience of about 1100 reactor-years. All outstanding safety issues, USIs, GIs, etc., would be subsumed by the program.

It wculd be made clear that the intent of the program wou'd be for this to be the end of new requirements for licensees. This would be changed only by the advent of important new infonnation or experience.

We note that the IPE program proposed by the NRC Staff already has been expanded well beyond the " search for outliers" concept.

In subsuming USI A-45, " Shutdown Decay Heat Removal Require-ments," into the IPE, for example, the Staff has taken a major step in the direction we are suggesting. Our proposal extends this to a more logical conclusion.

• Each licensee would be required to enduct a substantial and systematic risk analysis for their plant. We recomend that such an analysis would be a full scope PRA (at leastWe Level acknowledge

2) and include both external and internal initiators.

the difficulties inherent in making this an imediate require-ment. However, it should be possible to develop a phased ap-proach with the intent that within several years each plant would have been analyzed by state-of-the-art methods.

• Conclusions about results of the risk analysis and necessary changes in actual plant systems and procedures would be de-tennined by the licensee and reviewed by the NRC Staff through

The Honorable Lando W. Zech, Jr. May 10, 1988 the ISAP process. We believe the risk-based approach embodied in ISAP is the most logical means for resolving most safety issues.

The risk analysis used in the IPE for each plant will be available for use by the licensee and NRC Staff in their ISid evaluations.

We believe that the approach we have outlined above will provide the opportunity for a more integrated resolution of severe occident issues and other outstanding safety and licensing issues as well. We endorse current efforts on the part of the NRC Staff to formulate an inte-grated program.

Sincerely, W. Kerr Chairman

References:

1. U.S. Nuclear Regulatory Comission, NRR Generic Letter 88-02, dated January 20, 1988, " Integrated Safety Assessment Program II (ISAP II)."

Memorandum dated March 1,1988. from T. Speis (NRC) to D. Ross 2.

(NRC), et. al., "Comission Paper on Integrated Approach to Implementing the Comission's Policy on Severe Accidents" (Draft).

Memorandum dated April 1,1988, from T. Speis (NRC) to W. Kerr 3.

(ACRS), " Documentation Necessary for the Initiation of the Severe Accident Policy Implementation" (Draft Predecisional Attach-ments).

4 Draft SECY Paper (undated), Integrated Safety Assessment Program II (Predecisional Document), received April 26, 1988.

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• June 7, 1988 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20S55

SUBJECT:

INSERVICE INSPECTION OF BOILING WATER REACTOR PRESSURE VESSELS During the 338th meeting of the Advisory Committee on Reactor Safe-guards, June 2-4, 1988, we met with the NRC Staff to review the status of the inservice inspection program of boiling water reactor (BWR) pressure vessels. This matter was also considered during a meeting of the ACRS Subcommittee on Metal Components on May 26, 1988.

We were told that the NRC Staff is proposing a plan that would require the inspection of all accessible shell welds in BWR pressure vessels.

(Very few of these are now being inspected in the older plants due to lack of appropriate equipment and difficulty of access.) The plan would al sr> recomend a performance demonstration to show that the inspector and equipment are capable of detecting and sizing significant flaws in the reactor pressure vessel welds. Although we believe that catas-trophic failure of a BWR pressure vessel should continue to remain outside the design basis, recent experience demonstrates that flaws can grow from the coolant side of the pressure vessel into the steel pres-sure boundary. The proposed inspection program is desirable to maintain the appropriate defense-in-depth, and we therefore encourage its imple-mentation.

Sincerely, e W. Kerr i

Chairman WNS1 Attachment 6

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