ML20138H997

Transcript of 970502 Public Meeting in Rockville,Md W/Acrs. Pp 1-75.W/supporting Documentation

ML20138H997
Person / Time
Issue date:	05/02/1997
From:	NRC COMMISSION (OCM)
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References
REF-10CFR9.7 NUDOCS 9705070357
Download: ML20138H997 (173)

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NUCLEAR REGULATORY COMMISSION l

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Title:

MEETING WITH ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS) - PUBLIC MEETING Location: Rockville, Maryland i

Date: Friday, May 2,1997 Pages: 1 - 75 pPoW i i

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}A Washington, D.C.20005 m i x (202) 842-0034 50 970502 PT9.7 pop T

i DISCLAIMER This is an unofficial transcript of a meeting of the United States Nuclear Regulatory Commission held on May 2. 19_97 in the Commission's office at One The meeting was White Flint North, Rockville, Maryland.

This transcript open to public attendance and observation.

has not been reviewed, corrected or edited, and it may contain inaccuracies.

The transcript is intended solely for general informational purposes. As provided by 10 CFR 9.103, it is not part of the formal or informal record of decision of the matters discussed. Expressions of opinion in this transcript do not necessarily reflect final determination or beliefs. No pleading or other paper may be filed with the Commission in any proceeding as the result of, or addressed to, any statement or argument contained herein, except as the Commission may authorize.

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I 1 UNITED STATES OF AMERICA i

2 NUCLEAR REGULATORY COMMISSION l '3 ***

f 4 MEETING WITH ADVISORY COMMITTEE 5 ON REACTOR SAFEGUARDS (ACRS)

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7 PUBLIC MEETING l

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, 10 Nuclear Regulatory Commission 11 Commission Hearing Room l

12 11555 Rockville Pike 13 RockviAle, Maryland 14

15. Friday, May 2, 1997 i

16 17 The Commission met in open session, pursuant to 18 notice,-at 9:02 a.m., the Honorable SHIRLEY A. JACKSON, l 19 Chairman of the Commission, presiding.

20 21 COMMISSIONERS PRESENT:

22 SHIRLEY A. JACKSON, Chairman of the Commission 23 KENNETH C. ROGERS, Member of the Commission 24 EDWARD McGAFFIGAN, JR., Commissioner. .

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1 STAFF AND PRESENTERS SEATED AT THE COMMISSION TABLE:

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l 3 JOHN C. HOYLE, Secretary l 4 KAREN D. CYR, General Counsel l 5 WILLIAM. SHACK, ACRS I

j 6 JOHN BARTON, ACRS I

l l 7 MARIO FONTANA, ACRS 8 THOMAS KRESS, ACRS ,

9 ROBERT SEALE, ACRS 10 DANA POWERS, ACRS l I 11 GEORGE APOSTOLAKIS, ACRS l 12 DON MILLER, ACRS l '13 14 15 l 16 17 18 f 19 1

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1 PROCEEDINGS 2 [9:02 a.m.]

i I 3 CHAIRMAN JACKSON: Good morning,-ladies and -

4 gentlemen. l 5 It is a pleasure to once again meet with Dr. Seale ,

6 and members of the NRC's Advisory Committee on Reactor ,

t 7 Safeguards, who plan'to discuss a number of topics of 8 interest to the Commission at today's session.

9 Before I launch in, my colleagues apologize. They 1

10 are on travel and not able to be here. I 11 The ACRS provides advice to the Commission on the 12 safety of proposed and operating nuclear plants as well as 13 on safety-related policy matters, rules and regulations, I

14 elements of the NRC safety research program, prioritization, 15 resolution, implementation of generic issues and the use of 16 probabilistic risk assessment. The Commission is fortunate 17 to be able to draw upon views and experiences of this 18 selected and select group of technical experts as we try to 19 solve and address technical concerns in licensing and 20 regulation.

21 During today's briefing, we will cover the 22 following topics: Risk-informed performance-based 23 regulation and related matters, risk-based regulatory 24 acceptance criteria for plant-specific application of safety 25 goals, proposed regulatory approach associated with steam I

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l l 4 1 generator tube integrity, low power and shutdown operations 2 risk, status of ACRS review of the National Academy of 3 Science's National Research Council Phase II study report on 4 digital instrumentation and controls systems --

5 COMMISSIONER ROGERS: That's a mouthful.

6 CHAIRMAN JACKSON: Yes, it is.

7 Human performance program p.an and the ACRS report 8 to Congress on nuclear safety research and regulatory 9 reform.

10 Dr. Seale, my fellow commissioners and I welcome i l

11 you to this meeting and we anticipate another candid and 12 informative session with the committee and I understand that 13 if there is any briefing material, it has already been made 14 available.

15 So, unless my colleagues have any opening remarks, 16 please proceed. We have a full agenda.

17 DR. SEALE: Thank you, Chairman Jackson. We are 18 certainly happy to be here. We appreciate the opportunity 19 to convey to you some of our views on a first-hand basis.

20 We do have a full plate today and we hope we can get through 21 it in an expeditious but, more importantly, informative way.

22 So without further ado, I think we will get 23 started and Dr. Apostolakis will tell us about risk-informed 24 performance-based regulation and related matters.

25 DR. APOSTOLAKIS: Thank you, Bob.

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5 1 Well, you have received our letter, but I would 2 like to make some comments as an introduction to that letter 3 and the whole effort.

4 I think my fellow members agree with that, that-5 these regulatory guides, especially 1061, are a major or 6 significant achievement. Twenty-three years after draft 7 WASH-1400, we are finally using PRA. We are finally 8 recognizing that there is value to it. We stopped talking 9 about PRAs, good PRAs, bad. We are looking at specifics 10 now, what is modeled well, what is not modeled well.

11 The set of principles that are stated there, in my 12 opinion, are very good. They are the foundation of a new 13 regulatory philosophy. We finally recognize that sacred 14 cows such as defense in depth are not completely separate 15 from PRA, that one can see the lack of defense in depth, for 16 example, is reflected in some of the results of the PRA. So 17 I think this is really major progress and also we should 18 bear in mind we are talking about releasing these guides for 19 public comment. This is not the final version.

20 So, as far as I am concerned, the numbers that are 21 there, for example an incremental -- the small increases in 22 core damage frequency, they have to be 10 to the minus 6 or 23 whatever per year, these numbers will be scrutinized at the 24 next round, so we don't have to worry about it now. I think 25 the documents should be published because the industry is ANN RILEY & ASSOCIATES, LTD.

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6 1 very anxious to see some progress in these and they have not

• 2 seen anything yet, except for the viewgraphs that have been 3 used at various meetings.

4 So I don't worry too much about the numbers, j 5 except, of course, for the major numbers like the QHOs, 6 which are the Commission's policy. These numbers are not 7- subject to change but other numbers that are proposed in the 8 guides, I don't think we should worry about them.right now.

9 In fact, we will get feedback from the industry after we 10 release these guides from public comment.

11 I would like to come back to defense in depth and 12 safety margins. As I said, we made significant progress 13 there. I think we now understand better what the

• 14 relationship is between these two concepts and PRA.

15 CHAIRMAN JACKSON: Why don't you make such a 16 statement for the record as to what the relationship is, as 17 the committee sees it?

18 DR. APOSTOLAKIS: Well, basically, with regard to 19 defense in depth -- well, I think it also applies to safety 20 margins, the moment you try to talk about that relationship 21 you' realize that you have to consider what PRA models well 22 and what it models poorly and what it doesn't model at all.

23 Now, for things that are not in the PRA, first of 24 all, you have to find out why they are not there because 25 maybe they were considered and dismissed as intignificant.

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1 But, for these, it seems to me, these traditional approaches i

2 such as defense in depth and good engineering practices and 3 so on, then they can be applied the way we have been ,

4 applying them because they are not in the risk model.

5 But for other things that are in the risk model, 6 then I would look at the major contributors to risk. I 7 would look at the numbers, how high they are. I would look 8 at the uncertainties around these numbers and then I would 9 try to understand better these major contributors and ask 10 myself now do I have enough diversity here, do I have 11 sufficient number of barriers here. In other words, the 12 defense in depth idea but now I am doing it in a 13 quantitative way rather than relying on people's experience 14 which is not necessarily bad but this is better.

15 And then you can take it from there and go more -

16 deeply into it and so on but I think now we have a basis, a l

17 quantitative basis, in which we can implement this 18 philosophy. In fact, speaking of philosophy, it was my i 19 understanding that the second and third principles were 20 supposed to be stated as maintained the defense in depth 21 philosophy, not defense in depth. Because the concern was 22 that you can do a great probabilistic analysis and then i

23 somebody says, in the name of do"ense in depth, I don't like '

24 it. But the philosophy, I think, is a good idea that we 25 want to have multiple barriers, you don't want to rely on ANN RILEY & ASSOCIATES, LTD.

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\ 8 1 one single element in a minimal cut set, even though that 2 may have very low probability and so on.

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3 So the same thing goes for safety margin.

4 Somehow, the words philosophy were dropped and I guess we 5 will have to talk with the staff about it.

6 The first applications of this, of these guides, 7 will require team effort, in our opinion. We are not ready 8 to rely on a single reviewer to review their requests for 9 changes in the current licensing basis simply because this 10 is very new. And, again, it is not a new method. It is not ,

11 a new computer code; it is a new approach, it is a new 12 philosophy again. I guess I use that word a lot, 13 " philosophy," but I think it is important.

14 So it will require a team effort, a combination of 15 experts from various branches within the agency, until there 16 is a wide understanding, a common understanding as to how 17 this new approach will be implemented. Incidentally, in our 18 introductions with the staff, we tried to figure out whether j l

19 this was evolutionary or revolutionary. It was suggested it  !

20 was revolutionary with a small "r", then it was suggested it 21 was revolutionary with a Greek rho, so it is somewhere  !

22 between a revolution and an evolution.

23 Finally, I would like to state for the record that j 24 the Committee is extremely pleased with the cooperation that 25 the staff has shown in the last several months. We have had i

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9 1 excellent discussions with them and they were very willing 2 to listen to our ideas, debate with us and I found that a 1

3 pleasure, discussing technical issues at that level. I am '

i 4 sure my fellow committee members feel the same way.

i 5 Now, I didn't go into the details but you have the 6 letter and maybe if you have questions we can answer.  !

7 CHAIRMAN JACKSON: Maybe I will start out, ask you f 8 a few questions and then pass to my colleague.to my right 9 and then to my left. -

Can you tell us how have the pilot programs I 10 11 informed the development of the draft regulatory guidance 12 and standard review plan documents, or to what extent, and ,

13 what do you think has to happen at this stage? ,

14 DR. APOSTOLAKIS: well, the truth of the matter is  ;

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15 that the pilots were formulated before the basic approach 16 was formulated. It is probably due to administrative 17 reasons or whatever. I mean, the timing in my opinion was a >

18 bit unfortunate.  ;

I 19 CRAIRMAN JACKSON: The cart before the horse? '

l 20 DR. APOSTOLAKIS: Yes, exactly. I 21 CHAIRMAN JACKSON: And now I know from talking to 22 people that the pilots were put on hold in the last few 23 months, although the staff may disagree with me, because the i

24 staff was so busy preparing these documents and I think it '

25 is obvious that the preparation of these documents is not a t

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1 trivial matter.

2- So, now, on the other hand, I am reluctant to say 3 that the. pilots did not contribute anything to this because 4 the-pilot projects had already submitted requests and I am 5' sure the staff had read them, so that they had been 6 influenced by those but, in my opinion, that was the extent-7 'to which these documents were-influenced by the pilots. I B think the timing was unfortunate and that's why we recommend l 9 in our letter that new and innovative requests should be 10 solicited if possible by the Commission that will follow now i 11, this stated approach and we will see whether it works.

12 Now, I happen to have seen one or two of these i i

13 requests from the utilities, the current pilots. And, in my ,

14 opinion, it would not take much work to take what they have 15 done and cast it in this framework because the bulk of the  !

l 16 work has been done. They simply don't follow the boxes that  ;

i 17 we have in these in these because they were not aware of' 4 1

18 them. -t f

19 CHAIRMAN JACKSON: I think there is some review 20- going on relative to these documents and their discussions,  ;

21 I think. That's my understanding. )

l 22 DR. APOSTOLAKIS: So that is my impression.

23 CHAIRMAN JACKSON: What about the IPE reviews?

24 Were they -- did they inform the development of this draft 25 document?

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1 DR. APOSTOLAKIS: I believe they did. I believe 2 they did, especially when the staff proposed individual 3 numbers as to how high we want to go here, how high we want 4 to go there. I think they were influenced by the IPE 5 results. Also, quality of the analysis int he IPEs, I 6 think, was a major influence.

7 CHAIRMAN JACKSON: To what extent has the 8 Committee interacted with industry representatives on the 9 items documents and how would you characterize their views?

10 DR. APOSTOLAKIS: We have had presentations from 11 NEI and from South Texas Project representatives and I don't 12 remember now --

13 DR. POWERS: And Grand Gulf.

14 DR. APOSTOLAKIS: And Grand Gulf.

15 We found those interactions extremely useful 16 especially, as I recall, at the last subcommittee meeting we l 17 had two gentlemen from STP and it was a very intense 18 technical exchange and we felt that we benefitted a lot from 19 their perspective. So that was, I think, a very good 20 interaction.

21 CHAIRMAN JACKSON: Let me ask you one last sort of 22 linked set of questions. This is one of my favorite topics.

23 What role does uncertainty play in the 24 decisionmaking process? I mean, it seems to me that you 25 could have _wo plants with the same mean core damage ANN RILEY & ASSOCIATES, LTD.

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12 1 frequency that could lead to the same decision, although one '

I 2 could have an uncertainty of a factor of 10 and the other 3 uncertainty of a factor of 100. I mean, is this issue of 4 uncertainty and confidence intervals explicitly addressed or 5 doe s it need to be explicitly addressed? Do you think it 6 will be resolved in the public comment process? Give me 7 some sense.

8 Because the related question is whether the 9

proposed acceptance guidelines for core damage frequency and 10 large early release frequency would say, in effect, that no 11 increase in risk would be permitted. That is, can you 12 distinguish between 5, 10 to the minus four and 5.1, 10 to 13 the minus 4.

14 DR. APOSTOLAKIS: Well, I think how to handle 15 uncertainty was a major driver here because people are 16 uncomfortable with it. That is, in part, why I said this is 17 really a new philosophy.

18 For example, let me give you a few examples where 19 that concern influenced our interactions. Early on, one of 20 the early drafts of DG 1061, which is the main document, had 21 a figure or two figures that showed the core damage 22 frequency versus the allowed increase and there was a region 23 of acceptability, a region. There was a major discussion 24 regarding those so-called bright lines.

25 We argued very strongly that because of the ANN RILEY & ASSOCIATES, LTD.

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13 1 uncertainties this figure can be misleading, that if you are 2 a little bit to the left of the line it is acceptable, if 3 you are a little bit to the right it is not acceptable and, 4 as you say, who can tell? And we had all agreed that we 5 start working with the mean values but, of course, we have 6 in mind the whole spectrum of uncertainties and completeness 7 and so on.

8 So after a lot of debate -- because the text 9 itself was much more reasonable in my opinion in saying 10 look, the goal is 10 to the minus 4 per year for core damage 11 frequency, but, you know, there are many uncertainties, we 12 should recognize them, and so on -- so I was very unhappy 13 with the figures and I think finally the staff agreed that 14 we shouldn't have figures with those bright lines because of 15 the nature of this analysis.

16 Then there is a very good discussion in the guide 27 of the uncertainties in PRA, and again he comes back to 18 uncertainties in what is modeled, model uncertainty, 19 parameter uncertainty, things that are left out. So that's 20 very good progress too. Then the fact that the staff has 21 proposed this region of, let's say the goal is 10 to the 22 minus 4, then they say between 10 to the minus 5 and 10 to 23 the minus 4, there will be intense management or increased 24 management attention.

25 Now what does that mean? That means you ANN RILEY & ASSOCIATES, LTD.

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14 1 scrutinize the uncertainties. You look at it. You don't go 2 with the mean value only. You have to convince yourself 3 that what the request says makes sense. And I think at this 4 point in time this is very reasonable.

5 In other words, you cannot give, in my opinion, 6 specific rules and say this is what you do in that 7 situation, this is what you do in that situation, because we I 8 simply don't know. And that I think is one of the things l

9 that scares some people because now they will be responsible 1

10 for their actions. They will not have a guide or a table l l

11 that will say if and if and if, then. Now you have to use I 12 your judgment. For example, if you are in that region, do 13 you need seismic risk analysis? Do you need to worry about  ;

14 shutdown risk here? Do you need to worry about how well l

15 human error was modeled? These are questions that have to 16 be answered in the context of the specific request. But I 17 think three or four years from now we will know much more 18 about it, but right now it seems to me that's where it is.

19 CRAIRMAN JACKSON: Okay. Dr. Kress?

20 DR. KRESS: Thank you. I'm glad you asked that 21 question, because it's also one of my pet themes, how to 22 deal with uncertainties in the decision making process.

23 We're talking about having a criteria of what is an 24 acceptable risk. Now one could apply uncertainties 25 immediately in that in where we decide on what that level f

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1 is. We have essentially decided if it were to be the safety 2 goals that you use the mean, which is already a statement of 3 the uncertainties. You set the level, the .1 percent, at a 4 low enough value that it already accounts for your trouble 5 with the uncertainties. That's one area that you can deal 6- with uncertainties.

7 The other area is you're talking about dealing 8 with acceptable changes in risk, possibly. What is an 9 acceptable increase? And it's there where you might find 10 different levels of uncertainty because you have to evaluate 11 this particular thing and the ability to evaluate it is not 12 very good. So at that point is where I would think one 13 would talk about confidence levels. You talk about the 14 confidence level in your prediction of that delta risk for 15 that specific change, and I would have in my criteria that 16 you'd have to know that within some confidence levels. Now 17 I don't know what the appropriate choice for that would be.

18 CHAIRMAN JACKSON: Well, that was going to be my 19 last question on this topic, which is whether the choice of 20 confidence level is inherently a policy decision.

21 DR. KRESS: I think it is. I think it is. You 22 know, it's not something you can really technically say this 23 is what it ought to be. It's a decision, policy decision.

i 24 CHAIRMAN JACKSON: Right. Thank you.

I 25 Last question. Are these regulatory guidance ANN RILEY & ASSOCIATES, LTD.

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16 1 documents likely to have impacts on our regulatory analysis '

2 and rulemaking activities?

i 3 DR. APOSTOLAKIS: I think they will.

4 CHAIRMAN JACKSON: I mean, is there consistency l 5 with our regulatory analysis guidelines?

6 DR. APOSTOLAKIS: I would defer to one of my 7 colleagues who is more familiar with the regulatory )

8 analysis.

9 DR. KRESS: I could express an opinion.

10 DR. APOSTOLAKIS: Go ahead. I'm too new.

11 DR. KRESS: I think they are consistent.

12 CHAIRMAN JACKSON: Okay.

13 DR. KRESS: And they're consistent in several 14 respects. The regulatory analysis talk about substantial 15 changes and they talk about conformance with the safety i 16 goals and the process that they use in establishing the risk 17 and benefits or PRA's. I think they're consistent. There 18 may be some minor inconsistencies, but the philosophy is 19 essentially the same, and it wouldn't take much to make them 20 entirely consistent.

21 CHAIRMAN JACKSON: I don't usually do this. Mr.

22 Thadani, you would agree? You're nodding.

23 DR. THADANI: Yes, I would certainly agree. Our 24 intention was to -- Ashok Thadani, research. Our intention 25 was clearly to make sure that the approach we utilize here l

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1 is consistent with the Commission's safety goals of security 2 objectives as well as regulatory analyses that we use in our 3 .backfit decisions.

4 CHAIRMAN JACKSON: Okay. Thank you.

5 Commissioner Rogers.

6 COMMISSIONER ROGERS: Oh, yes, I wonder if -- I'd 7 like to come back to this uncertainty thing a little bit 8 later, but I wonder if you could just say a little bit more 9 about your comments with respect to graded quality assurance 10 where you felt that the staff was being unnecessarily timid 11 in their approach. Could you just sort of help me to 12 understand what you really have in mind there?

13 DR. APOSTOLAKIS: Well, it's a general impression 14 that's formed by reading the whole document. There is 15 extreme reluctance to categorize components or to declare 16 components or systems as belonging to the load safety 17 significant category. There is extreme reluctance to trust 18 or to believe that there is some information there. The 19 importance measures which the' industry is proposing. So on 20 top of that now we have the significant safety functions, 21 and it's not clear when you read the guide whether 22 everything that supports a safety function is itself of high 23 safety significance or not.

24 Anticipating your question I went back and I was 25 looking for a smoking gun. I couldn't find it. So it's a  ;

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I 18 1 general impression. And then of course what you.do with the 2 items that are in the low safety'significant category -- l 3 again there is significant disagreement between the staff I

4 and the industry as to how far'you go in relaxing the  :

5 requirements. f 6~ 'Now the other thing that puzzles me is that we are l 7 talking about something here whose value is not understood. I i

8 It is clearly stated in the guide that'we. don't have any  !

9 basis on which we can declare that QA makes the failure rate '

10 go down by a factor of 2 or 3 or 1-1/2 or the square root of 11 5. I don't know. We don't know what the benefit is, and 12 yet we're making such a big deal about moving things from ,

13 one category to another, as if, you know they will be i

14 completely inoperable if you put them in the low safety  !

15 significant category. And that I must say is really a very [

16 interesting and puzzling situation. I think it's another --

f 17 I think it's primarily tradition again, but people are so l 18 comfortable with QA that they feel very uncomfortable that f, 19 we will not'do these things to some components, _and the l 20 savings here -- l 21 CHAIRMAN JACKSON: So is your argument that ,

1 22 recategorization does not affect operability?

23 DR. APOSTOLAKIS: Oh, it might, but I don't know 24 by how much, and I don't think anybody does, and I asked 25 people, the. staff, is it a factor of 10? They say no. So I ANN RILEY & ASSOCIATES, LTD.

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19 1 don't know. We don't have any evidence. Maybe it would be 1

2 worthwhile to do something to try to understand that. How 3 much do we lose by recategorizing a system or a component?

4 COMMISSIONER ROGERS: Well, I think that that 5 relates to the concern which I've been hearing from some 6 industry quarters that they don't really see any benefits 7 from the use of PRA yet from NRC's regulatory posture other 8 than it's another way of looking at things and it certainly 9 is useful in understanding the plant, but in terms of 10 regulatory relief -- I kind of hate that word, but I don't 11 know what I've -- I haven't got a better phrase, but I think 12 we all know what we're talking about -- some modification on 13 the basis of reclassifying requirements on the basis of 14 greater knowledge of their safety significance when they may 15 have had a perfectly reasonable historical origin that 16 seemed like a good idea at the time, but the time was a long 17 time ago, and now we know a lot more, but there doesn't seem 18 to be much action in that direction.

19 I wonder if you have any thoughts as to how to 20 approach this in a systematic way because obviously you .

21 can't do everything at once, but would there be some area of 22 application of PRA that is so sound, so incontrovertible, 23 that one could simply use PRA to take some steps on removing 24 or changing regulatory requirements which clearly on a 25 quantitative and even expert judgment point of view based on ,

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20 1 historical experience really just simply don't make any

• l 2 sense anymore. I know we've talked about things like 3 limiting conditions for operations that have come out of I 4 some PRA studies, but I don't know if we've changed anything 5 as a result of that. So I wonder if you have some thoughts 6 there, because this is really a very important area. l 7 The other related observation, I think your 8 comments with respect to the uncertainties are extremely  ;

9 interesting -- I have to say a bit disappointing, however, 10 because I think that one of the -- not that I disagree with 11 them -- but that one of the advantages to moving to PRA I ,

1 12 think is to allow the NRC to be able to point to PRA )

l 13 analysis as a more objective set of measures for decision I l

i 14 making than what have been used in the past when we've used 15 things like expert judgments and good engineering practice, 16 which we feel very comfortable about but are hard to explain 17 sometimes to the public.

18 DR. APOSTOLAKIS: Well I think, starting with the 19 uncertainty, I think it is a more objective way. The 20 problem is we cannot use decision theory as is. Decision 21 theory tells us we should work with mean values. I think in 22 our industry we're using mean values as a first step. The 23 degree of uncertainty, the level of uncertainty, is very 24 critical to us, because if the uncertainty is very large, we 25 may change the decision problem and say well, we have to l

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1 understand this better. If we start a new research project 2 instead of making the original decision.

3 So we are not really following a well-established 4 mathematical theory, and that's what the problem is, that we 5 cannot work only with mean values, because as the Chairman 6 said, you can have examples where the means are the same, 7 but the spread is very different, and then of course you 8 can't tolerate that given the hazards we're dealing with.

9 So I am not sure that at this point we can go beyond what I 10 said. In fact I would be very reluctant to accept anything 11 more prescriptive in terms of 95th percentiles and so on. I 12 think it's too soon. I think too few people understand 13 these things, and again, I don't want to be as prescriptive 14 as in the current system in the new domain, so I think it's 15 something to think about, but I really think that issue has 16 to be resolved by the reviewers, and I frankly think that's 17 why people don't like PRA. They have to now make decisions, 18 not just follow rules, and that's why the first several 19 cases we have to have a team doing the review, so you have 20 the right expertise there to make people feel comfortable.

21 COMMISSIONER ROGERS: Thank you.

22 DR. SEALE: If I might make a comment, I think in

! 23 terms of the appropriate, perhaps, applications that might l 24 demonstrate the validity and value -- or let's say the value l

l 25 of PRA, the industry has already voted once, in a sense, in l

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1 22  !

1 1

that they identified the pilot topics as being areas in

• 2 which they felt the expenditures they were making were l

3 potentially worthwhile.

{

4 l

I think we ought to, as we move along, and as  :

1 5

George mentioned earlier, those submittals ire in many cases 1 6

in extraordinarily good shape, require some recasting to 7 satisfy the guidelines, but that we ought to try to handle 8

those as soon as we can in a reasonably expeditious way.

9 I think also the invitation to the industry to 10 propose other candidate areas should be encouraged. Again, 11 I think we have to recognize that in some cases, 12 particularly where I won't say regulatory but financial 13 relief is a candidate, that we have to expect to be in some 14 respects in a reactive position in that the industry is much 15 better able to identify the loads that they consider to be 16 inappropriate or unduly onerous. 1 17 CHAIRMAN JACKSON: And my understanding is that 18 there is -- that there is work relative to how PRA and the 19 guidance that would come out of these documents might be 20 used in the areas not only of graded QA but with respect to 21 tech specs, technical specifications, inservice testing, and 22 inservice inspections, isn't that correct?

23 DR. SEALE: Yes.

24 DR. APOSTOLAKIS: Yes.

25 DR. SEALE: And those are the things that could i.

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23 1 very well be pushed forward.

2 CHAIRMAN JACKSON: Right, and in fact there is i l

3 activity as far as I understand in moving along that line.

4 DR. SEALE: I think some of the industry people 5 are now kind of waiting for the shoe to drop'on submitting i

6 1061 because they haven't seen it.

7 CHAIRMAN JACKSON: Okay, and I think we are going 8 to be getting a briefing from the Staff, maybe it is next  !

9 month or later this month on the PRA implementation plan --

10 maybe it is next week'-- and I think we are going to get a 1

11 complete update in those particular areas. l 12 DR. APOSTOLAKIS: I don't believe I answered your  !

13 first question though, Commissioner Rogers, regarding the l 14 ~ ' quality assurance.

i 15 At one of our subcommittee meetings we had a 16 presentation from the South Texas Project folks and they '

17 -stated that if they are allowed to do what they propose,  !

18 they would be saving about $1,300,000 a year, just from i

19 that.  ;

20 Now what do you do about it? I mean there is 21 .obviously disagreement between the staff and the industry on .

l- 1 22 this and us. We received a letter from the EDO that states 23 that the Reg Guide now has a new version that accommodates

-24 some of our concerns but we have not seen it yet.

25 I don't know why we don't go ahead and implement ANN RILEY & ASSOCIATES, LTD.

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i 24 1 one of these proposals from the industry. In 1061 there is j 2 a clear box there that says that in the decision-making 4

3 process that the licensee should propose a monitoring l

4 program and we have integrated decision-making and 5 everything. Okay. We don't even know what the benefit of 6 QA is. Let's implement the program. Let's have a good f I

7 monitoring program there and see three, four, five years 8 down the line whether the lack of this quality assurance i 9 which is a result for the high safety significant components f i

10 really inskes any difference to the other components, and if  ;

11 it doesn't, then that's fine. We learned a lot.

12 I mean it is not something that is cataclysmic, in 13 my opinion. I don't understand what the big deal is.

14 CHAIRMAN JACKSON: Okay. Commissioner McGaffigan. 3 15 COMMISSIONER McGAFFIGAN: I'm sorry to extend this j i

16 part of it --  :

l 17 CHAIRMAN JACKSON: Please. j 18 COMMISSIONER McGAFFIGAN: -- but on the issue of 19 benefits to the industry of proceeding down this path in 20 graded QA, inservice inspection, inservice testing, how 21 widespread are the benefits going to be in the sense that 22 how much of industry has good enough.IPEs, PRAs that they 23 will be able to take advantage of whatever relief is 24 implicit in these initiatives, beyond the few pilot plants 25 that -- the South Texases, the Palo Verdes, whatever, which 7dDJ RILEY & ASSOCIATES, LTD.

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l 25 l 1

apparently have good PRAs?

1 1

2 People have said to me, even PRA advocates, that 3 some of the IPEs are sort of junky, and how widespread will 4 the benefits be?

5 DR. APOSTOLAKIS: Well, fi.rst1of all, I will 6 answer your question, Commissioner, but the focus of our review of these documents was not, you know, how widespread .

7 8 the benefits will be.

9 We looked at --

t 10 COMMISSIONER McGAFFIGAN: Right, I understand.

11 DR. APOSTOLAKIS: -- at philosophy and safety and ,

j 12 unnecessary burden, and so on, l

13 Now I believe, and again I haven't done any  ;

14 scientific polling on this, but the number of utilities that 15 will benefit in the very near future from these guides is r

16 small because it takes a certain sophistication in the PRA 17 area to be able to formulate the request the right way and ,

l 18 use the right terms and so on, and have the right approach, 19 but I think it will spread very quickly.

20 It will spread very quickly. One of the problems, [

21- I think, is that a lot of the decision-makers in the 22 industry either do not know at all or are not convinced that l l

23 PRA will be useful to them. I-happened to organize a course '

24 at MIT last January for mid-level managers at utilities and 25 plants that make decisions, and the subject was how to use j ANN RILEY & ASSOCIATES, LTD. ,

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-1 PRA to make decisions, and it was interesting to see how '

2 most of them had no idea what PRA could do with them, and 3- then towards the end of the course they could see -- you 4 know, for instance, the importance measures, how they can be 5 used to help them with problems they are facing right now.

6 That doesn't mean that this fellow now is ready to 7 do it tomorrow, because.he has to learn and he has to have 8 the organization and so on, so at the beginning it is my 9 opinion, and maybe others disagree, there will be a small 10 number, but if the dollar numbers we are hearing from those

! -11 expert utilities right now are true, then it seems to me 12 that word will spread very quickly, very quickly.

13 COMMISSIONER McGAFFIGAN: Does anybody else -- let 14 me ask a question.

15 As I understand it, when we implement these 16 various initiatives, we are doing it through license 17 amendments? Is that correct? So we don't get into 5059 18 space or -- the fundamental issue, you all are urging us to 19 allow for small changes, and that is the direction we may be

( 20 going if the legal analysis we have asked for buttresses 21 that, but in.5059 space, the plain reading of 5059, if we 22 have to implement any of this through unreviewed safety --

23. the heart of it is in unreviewed safety questions, is any I 24 increase in the probability may result in an increase in l

25 . probability. That is the Staff's view. We have it out

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27 l' there.

2 Do you have any thoughts as to if you want to go 3 down this path in making broader use of PRA whether there 4 has to be changes made in 5059 and the definition of an i

5 unreviewed safety question in 5059?

6 DR. APOSTOLAKIS: I would let colleagues that have >

7 been on the committee longer than me --

8 [ Laughter.)

9 DR. APOSTOLAKIS: -- answer this question. I'm 10 sorry.

11 CHAIRMAN JACKSON: Dr. KrEas?

12 DR. KRESS: Since I am the senior member, which 13 really seems strange to me, I can give you an opinion.

14 My opinion is that these are parallel paths, that 15 the 5059 is not affected by this process at all, and this 16 process we are talking about.is in the form of a change to 17 the licensing basis --

18 COMMISSIONER McGAFFIGAN: Right.

19 DR. KRESS: -- and you continue with the 5059 20 process as it is. You don't need to change the rules in it 21 or what constitutes an unreviewed safety question.

22 You keep that all the same, and that. allows the 23 plants to continue making those changes which are allowed 24 within that route.

25 COMMISSIONER McGAFFIGAN: Within the Staff's --

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28 1 within the reading of the rule as it has been currently 2 propagated? .

3 DR. KRESS: Yes.

4 CHAIRMAN JACKSON: Let me make sure I understand 5 something, and I don't want to be jumping in here but I am 6 jumping in here.

7 You seem to be saying, and I don't want to put 8 words into your mouth, that there are classes of changes 9 that would be within the scope of 5059 that can be left 10 alone.

11 DR. KRESS: That's right.

12 CHAIRMAN JACKSON: But there are other classes of 13 changes that may involve some increases in risks within some 14 margins that should then come to the Commission --

15 DR. KRESS: --

through this other process --

16 CHAIRMAN JACKSON: -- come to the Staff --

17 DR. KRESS: That's what I am saying.

18 CHAIRMAN JACKSON: -- and would be governed then 19 by and guided by --

20 DR. SEALE: License amendments.

21 CHAIRMAN JACKSON: -- license amendments that L2 would also involve the PRA analysis.

23 DR. KRESS: Yes, exactly.

24 CHAIRMAN JACKSON: I understand. So you are 25 saying that in fact there is a possibility to do a ANN RILEY & ASSOCIATES, LTD.

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1 29 1 bifurcation, namely if your plant change can satisfy the 2 reading of 5059, do it --

3 DR. KRESS: By all means --

4 CHAIRMAN JACKSON: -- go ahead and do it and you 5 don't have to come in for a license amendment.

6 DR. KRESS: Absolutely.

7 CHAIRMAN JACKSON: But if in fact it may involve a 8 change in risk, you bring it in --

9 DR. KRESS: Through the other process.

10 CHAIRMAN JACKSON: -- through the other process,

11 the more formalized, to which this kind of analysis can be 12 applied. That's very interesting.

13 COMMISSIONER McGAFFIGAN: And just to follow up on t

14 that, the benefit of the reg guides then for somebody coming 15 in under the formal process will be that they will have 16 certainty in advance as to how we are going to look at the l

l 17 change.

18 DR. KRESS: That's exactly right, yes.

19 COMMISSIONER McGAFFIGAN: So that the advan'es 20 being made through the pilot programs and the reg guides is 21 to define the parameters under which we will typically look 22 at an amendment that involves an unreviewed safety question 23 and involves potential changes in risk.

24 DR. KRESS: That's a very good way to look at it, 25 yes.

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30 1 1 COMMISSIONER McGAFFIGAN: Okay. '

i 2 CHAIRMAN JACKSON: That is very interesting. That 3 is a useful clarification, and my understanding is -- I i

4 think we are going to need to move on -- that in next week's  !

5 meeting on the PRA implementation plan we are going to hear f 6 more specifically, Mr. Thadani, about where the various 7 pilots stand?

8 MR. THADANI: Yes.  ;

9 CHAIRMAN JACKSON: Okay, so I think that with that  !

10 we will, if we may, move on to our next topic, and I think  ;

11- it is related. Dr. Kress, f 12 DR. KRESS: That's true. It is related. It is a .

13 sort of a sub-area within that whole larger, broader area.  ;

14 The topic is about acceptable risk criteria, i

15 safety goals, and adequate protection and interrelationships i 16 of those things.

I 17 We see in this process of the 1061 that it was

)

18 necessary to come up with some quantified level that we 19 would call an acceptable risk.

20 And it seems to be a necessary thing if you are 21 going to really have this type of process which would be )

i 22 risk informed. But our body of regulations and the way they j 23 have developed and evolved over time is they are rooted in 24 the general design criteria and the design basis accident 25 concepts and the philosophy of defense in depth. And the ANN RILEY & ASSOCIATES, LTD.

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31 !

l' 1 presumption is if we do all of that correctly, you will end f l  !

2 up with a plant that provides adequate protection which, to t

l 3 me, really boils doen to adequate protection as we now know  :

l L

4 it is compliance with all the rules and regulations and l  !

l 5 commitments. It is not a quantified level of risk. l l 6 Well, that is a concept that has served us well, l l

l 7 has worked very well. I think it has resulted in plants ,

8 that do provide adequate protection but it is not a very I 9 useful thing in a risk-informed concept like we are talking 10 about now. You really do need to quantify this thing we  ;

11 call adequate protection or acceptable risk. I will use  ;

i l 12 those interchangeably.

13 The safety goals are an expression of what we feel  ;

14 like is how safe is safe enough. They are posed in risk 15 terms and it was our opinion that one has two options. They i

16 could decide if you want to quantify what we call an -

17 acceptable level of risk, one could just automatically 18 select safety goals because they have already been an i 19 expression of what we say is how safe is safe enough. Or I 20 one could try to quantify what we mean by adequate 21 protection. That is a difficult process to quantify that.

22 It is our feeling that it is a risk level that is  !

23 higher than the safety goals and the reason we say that is 24 that since any plant out there that is licensed and 25 operating by definition meets adequate protection standards.

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32 1 Then the plant that has the highest level of risk puts a 2 bound on that.

3 We think there are plants that are both above and 4 below the safety goals. So it is our opinion that adequate 5 protection, if it were to be quantified some way, is above 6 the safety goals.

7 Our choice, our recommendation for picking a value 8 to use in this new process of risk informed for acceptable 9 risk favor is the safety goals, which get you below adequate I 10 protection level and is a quantifiable level that we can 11 deal with.

12 CHAIRMAN JACKSON: What would that then do to 13 plants that are above that?

14 DR. KRESS: At the present time we are not talking i

15 about enforcement.

16 CHAIRMAN JACKSON: So you are talking about as a 17 pattern.

18 DR. KRESS: We are talking about decisions on the 19 acceptable changes to the licensee basis. Now, I think in 20 the long run, one would like to view the safety goals as a 21 replacement for adequate protection and one would like, in 22 the long run, to actually enforce that. I think we have a 23 great deal of difficulty with that because of backfit rules 24 and --

25 CHAIRMAN JACKSON: But let me make sure I l

l '.

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' i 1 understand where you are. Again, I am going to paraphrase 2 it and if I am doing it wrong, you tell me.

3 DR. KRESS: You do it much better than I do. ,

l I

4 CHAIRMAN JACKSON: I doubt that.

5 You were saying at this stage of the game, de 6 facto, the fact that we are allowing plants, the universe of 7 plants to operate, means that we have said they provide 8 adequate protection that is adequate.

9 DR. KRESS: Clearly.

10 CHAIRMAN JACKSON: So then if one wants to look at 11 the safety goals and you use it for decisions on what 12 constitutes acceptable changes to the licensing basis, what 13 you are really then saying is if the safety goal is where we 14 want to place that threshold that while there are plants 15 that are currently allowed to operate that are above it, 16 that if they wanted to change their risk profile they would 17 be more constrained than plants that are currently below it?

18 DR. KRESS: Very good.

19 CHAIRMAN JACKSON: Is that what your basic point 20 is?

21 DR. KRESS: Absolutely.

22 CHAIRMAN JACKSON: I want to be sure I understand.

23 DR. KRESS: Very well put.

24 So we are safe, I think, in using the safety goals 25 as an acceptable risk criteria but these are expressed in ANN RILEY & ASSOCIATES, LTD.

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I j l

34  :

1- terms of prompt fatalities and latent cancer deaths which 2 does require a level three PRA. There is no way around 3 that. You cannot determine those things without level three -

4 PRAs. To use a level three PRA in this concept of risk-5 informed acceptable changes is a bit awkward, to say the 6 least. It is not very -- it doesn't really focus one's 7 attention on the plant features and the things that are 8 safety significant. So it would be much better if one could i 9 have more tiered criteria, such as the core damage frequency 10 and the conditional containment failure probability. But 11 still be within the confines of the QHOs.

1 12 In our December meeting, I said that was entirely '

13 a possible thing to do, to derive these lower tier criteria l

14 directly from the QHOs. Well, you pinned me down and said, l l

15 all right, when can we see that and, being the eternal 16 optimist that I am, I say within a few weeks, I think is 17 what I said.

18 Well, we are now, with our recent letter of April 19 11, we are providing that to you. I must say, though, in my 20 defense that I did have it ready within a couple of weeks.

21 We are, however, a committee.

22 CHAIRMAN JACKSON: You just had to propagate it?

23 DR. KRESS: That's right.

24 CHAIRMAN JACKSON: So if the values are derived )

25 from the prompt fatality QHOs, how much would the core l

l 1

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35 1- damage frequency or large early release frequencies change 2 from site to site? Do you know?-

3 DR. KRESS: I don't know because I haven't done 4 that yet, that exercise.

5 CHAIRMAN JACKSON: Can you do that for us?

6 DR. KRESS: I can, yes.

7 In the attachments to our letter, we provide a 8 technically sound, rigorous way to do that.

9 CHAIRMAN JACKSON: To do that? Okay.

10 DR. KRESS: In fact, as you could understand, it 11 does have to make use of level three information but, 12 fortunately,- there is enough level three information out 13 there to be able to do it without having to go back and do 14 level three for every plant.

15 In fact, one of the attachments was a very nice 16 analysis made by our senior fellow, Rick Sherry, which gives 17 a way to estimate the level three consequences based on 18 site-specific characteristics, which is a very nice piece of 19 work. That alone with the process I recommend for deriving 20 the lower tier criteria from the safety goals should be very 21 useful to the staff in this whole process of determining 22 risk acceptance criteria in terms of core damage frequency 23 and LERF or conditional containment failure probability.

24 CEAIRMAN JACKSON: Commissioner Rogers?

l 25 COMMISSIONER ROGERS: Well, this has been a topic l

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I 36 1 of real interest here for a long time of how do you deal

• l l 2 with the fact that the safety goals are founded on level i

3 three PRA results and there are different circumstances at 1

! 4 each plant that have absolutely nothing to do with the plant j 5 design or operation. They are where it is and what the l 1

l 6 meteorological conditions are nearby and all this sort of l 7 thing.  !

8 I have raised the question in the past, and it has

{

i l 9 always sort of led down a path that goes to nowhere and that i

10 is could one have a kind of standard location, population i 11 distribution and so on and so forth of some sort that more I 12 or less bounds whatever exists with our current level, our I

13 current plants, and then just say that is the one you are 14 going to plug in when you go to look at effects of changes l l

15 in anything else in the plant. You know, if you want to l l

16 then take the next step of applying those, the effects to j

17 health effects, that then you would have a standard 18 population distribution, so on and so forth, that you would 19 always balance it against to see what the effects were.

20 Apparently, somehow or another, that never seemed 21 to be doable. I think we have talked about it occasionally 22 in the past.

23 In effect, it seems to me that when you go to 24 surrogates for the health effects, you really are doing 25 something like that, aren't you? Because you are not l

l l

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37 1 looking at the health effects, you are going to just ignore ,

2 them and therefore you are going to create something else 3 which stops short of the health effects but you are willing 4 to accept and, in a sense, it seems to me philosophically 5 that is about the same thing.

6 I wonder, it doesn't have a quantitative health i 7 effects -- you don't get a quantitative health effects l 8 number out of it, you stop short of that. Wouldn't it be -

9 still nice to be able to do that? ,

10 DR. KRESS: It would be. But let me -- let me 11 tell'you about two attachments. [

12 They actually do quantify the health effects. It ,

t 13 is a way to do it on a simpler -- it is approximate but it. l 14 is a very good approximation. It makes use --  ;

15 COMMISSIONER ROGERS: But it still would be site  ;

16 specific?

17 DR. KRESS: It would be site specific. It makes 18 use of site specific population parameters, site specific t 19 meteorology. And so it is a way to do a level three in a 20 much simpler, much, much simpler. And you can back out of t

21 that in site specific values that you would use for a LERF 22 or a core damage frequency and conditional containment f

23 failure probability. That would be site specific. You 24 would have a different value for each site to meet the l i

25 safety goals.

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38 1 I am not sure, that would be one way to go. It 2 may be a little awkward because you have a different set of 3 criteria for each site. Another way to go would be to take 4 the site that bounds these two things and use that as your 5 criteria and you know you are safe with all others then.

6 The staff has chosen to do this latter at the 7 moment, to take a bounding. While the problem there was 8 they chose a number of sites and evaluated them to get this 9 bound, they didn't take all sites. I am not real sure --

10 CHAIRMAN JACKSON: They told as a complete bound.

1 11 DR. KRESS: Yes, I am sure it is a complete bound.  !

1 12 I am not quite sure that their process of backing into the l l

13 CDF and the LERF was as rigorous as the one we are 14 recommending in our attachment. But they did a good job 15 with that. They did it right. Their option right now is to 16 use a bound, which I think is good because it gives you one 17 set of criteria and you don't have to deal with each 18 individual site that way.

19 CHAIRMAN JACKSON: Dr. Apostolakis.

20 DR. APOSTOLAKIS: Yes, I think it is important in i 1

21 this discussion to bear in mind that the committee is on l

22 record recommending that the core damage frequency be 23 elevated to a fundamental objective level and be independent l 24 of a site, independent of everything else. And the value of 1

25 10 to the minus 4 for the reactor year we thought was a .

1 1

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l

1 39 1 reasonable number because it is not just the health effects 2 of the accident that are important but the fact that you 3 have had something, a serious thing, is very important. So i 4 that is something we want to prevent. 1 5 In fact, if you work backwards, we are talking  ;

6 about LERF here, you end up in some sites with a core 7- damage -- acceptable core' damage frequency which is higher 8 than 10 to the minus 4 per year and we felt, as a committee, 9 that we don't want to live with that.

10- So the whole discussion really concerns LERF only.

11 CRAIRMAN JACKSON: Commissioner McGaffigan?

12 COMMISSIONER McGAFFIGAN: Pardon my skepticism 13 .about this stuff but on PRA, I have had discussions with 14 various folks including at the reg info conference.

15 CHAIRMAN JACKSON: Commissioner McGaffigan and I ,

16 are going to write a PRA paper.  ;

17 COMMISSIONER McGAFFIGAN: Yeah.

18 People tell me, and you correct me if I am wrong, 19 that --

or they are wrong, that PRAs can be pretty good at- i 20 looking at incremental changes, when you make a change, but, 21 you know, people tell me not to believe core damage 22 frequency numbers to better than an order of magnitude and 23 sometimes people correct me to two orders of magnitude.

24 So when you are talking about, as I say, I 25 understand that they may be very good at looking at  ;

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40 l 1 incremental changes but given the fact that you at the ~

2 outset, Dr. Apostolakis, said that there are things that 3 they model well, don't model at all, why should I believe  ;

4 this stuff.when we start talking about them as if you can >

5 calculate to 1.33 times 10 to the minus fourth?

6 CHAIRMAN JACKSON: That was my whole point about 7 uncertainty.

8 COMMISSIONER McGAFFIGAN: Uncertainty. No, I 9 agree.

10 DR. KRESS: Go ahead, George.. You were just 11 asked.

12 DR. APOSTOLAKIS: I think this is the -- we have 13 to get.away from statements like PRA is good, PRA is no 14 good, PRA does this, PRA doesn't do that. PRA deals with 15 the whole plant, it is not just a computer code doing one 16 thing.

17 Certain things PRA does very well. In fact, the 18 level one PRAs are pretty, good. They capture a lot of 19 important things so I would trust them, you know, when I 20 make decisions, depending on the decision.

21 I think we should talk about specifics. If we 22 talk about, say, human error and human actions recovery and 23 so on, then I would be a little more skeptical. Maybe I can 24 bound the number but I wouldn't really believe a 25 distribution that somebody gives me right now.

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41 1 Then this issue of design errors, organizational 2 issues and so on. But, in my opinion, just because I don't 3 model the organizational plant, does not reject the whole 4 approach.

5 So that's why it's really important to understand 6 what is modelled, what is modelled well and what is not 7 modelled, and then depending on the context of the decision, 8 you know, make a decision.

9 You know, there may be decisions where what you 10 are saying is absolutely right. I don't believe the numbers 11 because this affects something that is not modelled there, 12 but I believe the industry also believes this. We can make 13 very good decisions at the Level 1 PRA.

14 COMMISSIONER McGAFFIGAN: But if I just look at 15 core damage frequency --

16 DR. APOSTOLAKIS: Yes.

17 COMMISSIONER McGAFFIGAN: -- when I get a number 18 for a plant on core damage frequency, to what order of 19 magnitude should I --

is that --

20 CHAIRMAN JACKSON: It depends on the model.

21 COMMISSIONER McGAFFIGAN: Should I assume that 22 number is correct?

23 CHAIRMAN JACKSON: It depends on the model.

24 COMMISSIONER McGAFFIGAN: Okay.

25 DR. APOSTOLAKIS: I think it depends on who did it ANN RILEY & ASSOCIATES, LTD.

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l l

42 1 and also it's not really a number. '

I mean they have to give 2 you a distribution.

3 COMMISSIONER McGAFFIGAN: A distribution, yes.

4 DR. APOSTOLAKIS: A distribution. Now some of the 5 better PRAs -- it's really hard for me to see how the l

6 diUtribution or the upper end of it would really shift too 7 mudh ;o higher values because we have missed something.  !

8 We have been doing this now for over 20 years and  !

l 9 I don't think that we have found things like in the early 10 days, of course, the reactor safety study dismissed external 1

11 events, then the industry came back with the Zion, Indian 12 Point PRAs and said, no, fires and earthquakes may be i

13 significant contributors. l l

14 You don't see that anymore. You don't see these i 15 quantum leaps anymore. Now, you know, we are sharpening the 1

16 pencil here and there -- '

17 COMMISSIONER McGAFFIGAN: See, the thing that j 18 strikes me, at least some have told me that when you make an 1

19 incremental change you can understand the effect of the 20 incremental change even if the whole distribution may be off 21 a bit because perhaps human performance is going to be the i

22 same whatever -- you know, whatever test you are going to do 23 or whatever other change you are making in the plant, so i 24 differences are oftentimes easier than knowing the whole 25 curve.

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43 1 Is that not correct?

I 2 DR. APOSTOLAKIS: I don't think there is a 3 correct --

4 CHAIRMAN JACKSON: Could I address --

5 DR. APOSTOLAKIS: Let me just say what I feel 6 about it.

• 7 CHAIRMAN JACKSON: Okay.

8 DR. APOSTOLAKIS: I have never believed that that 9 was a rational. approach. That was my personal opinion. I 10 think the absolute number -- ,

11 CHAIRMAN JACKSON: Dr. Kress's comment is going to 12 be the last word because we are not going to be able to get ,

13 through the agenda here.

14 DR. KRESS: I would like to express an opinion on 15 this delta risk versus the bottom line.

16 A PRA basically integrates the risk contributions 17 from a lot of things. '

18 CRAIRMAN JACKSON: That's right.

l 19 DR. KRESS: It adds them up and if you could take >

20 the derivative of that integral, you would have a set of 21 partials added together due to each of these contributions.

22 Those partials are -- you can better define those i

23 partials. You can narrow down the uncertainties in each of 24 those partials. The uncertainties in the sum of all of them .

25 get very large.  !

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44 1 Some of the partials are different than others, so 2 it depends on the nature of what increment you are talking 3 about, but in general the incremental risk that you 4 determine due to the change is much more precisely known 5 than the bottom line, and that you can take as a given, and 6 it is easier to deal with those, but they still have 7 uncertainties in them and it will be a variable uncertainty 8 depending on which type of increment you are talking about.

9 CHAIRMAN JACKSON: Dr. Seale.

10 DR. SEALE: Thank you.

11 The next topic is the proposed regulatory approach 12 associated with the steam generator integrity issue.

13 I think I will try to expedite this a little bit 14 and see if we can get a little bit back. I do this in part 15 because we still haven't heard the final word from the Staff  !

16 on what they are going to come down with. We have a pretty 17 good idea of what they are going to have on that issue.

l 18 In any event, I do have to confess that back in l

19 1994, which shows the time constants on some of these l

20 things, we were a party to the decision that, or at least we 1

21 concurred in the decision to go to rulemaking on the issue 22 of steam generator tube degradation.

23 In the interval we have had numerous discussions l 24 with some of the Staff on some of the details in developing l

l 25 their approach to those issues in much the similar manner to i ,

J i

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45 1 Dr. Apostolakis's earlier reference to the work in the PRA I 2 area.

3 As a result, when in November of this year, of j 1

4 this last year, we finally got a look at the proposed rule )

5 and the associated Reg Guides, it was not a complete shock 6 to anyone that we had some serious reservations about some i i

7 of the things that were there, and all of them really 8 revolved around the problem that there was an inability to- -

9 identify a risk evaluation methodology that wold allow you 10 to take test data and come up with an assessment of risk due 11 to indicated degradations in tube integrity that would allow  !

12 you to justify continuing those tubes in service rather.than 13 going to the plugging strategy which has been the classical 14 way of handling that problem, j 15 There were specifics that went along with that  ;

16 difficulty, that is -- that grew out of it, but perhaps the 17 most significant thing was that the rule wound up or the 18 proposed rule wound up being an admittedly performance based t

19 regulation but it had very little in the way of risk i 20 objectives or risk information in helping or in justifying 21 those performance --  ;

22 CHAIRMAN JACKSON: So in that sense it diverges 23 from what the approach is in the -- ,

24 DR. SEALE: Sure. ,

25 CHAIRMAN JACKSON: -- relative to what we have i

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46 ,

e 1 just been talking about with ISI and IST -- ,

2 DR. SEALE: Right. ,

3 CHAIRMAN JACKSON: -- and so forth.

4 DR. SEALE: Exactly. Now there were a few other l 5 things there but we also make the point that there was an 6 outstanding generic issue and a differing professional 7 opinion that had to be cleaned up in this process as well, i 8 and I won't go into all of the details there. ,

9 But then in January we got a -- we sent a letter i

10 to the EDO in which we reiterated our concerns that we had j 11 expressed in our November letter and also brought up a few )

12 specific issues that members had identified having to do 13 with things like the risk due to thermally-induced tube 1

1 14 failure and severe accidents.

l l

15 There you get into severe accident space when you j 16 are supposedly more interested in -- or limited to design 17 basis accident considerations.

18 The Staff was then asked to go back and look at  ;

19 those issues in coming up with -- or to consider them in  !

i 20 coming up with their rule.

l 21 We met again with the Staff in March, and they 22 outlined to us what they proposed to be their approach, l

23 which would be to look at an alternate way of doing things, 24 basically to not go to rulemaking but to go back to 25 essentially the previous approach with some enhanced --

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47 l .

I well, I'm sorry, to use the current regulations and then j 2 also recommend a PRA implementation plan as a framework for ,

j 3 coming up with any alternate proposals for regulating the j 4 steam generator tube issues.  !

r 5 One of the things that we notices was that the --

6 or we commented on was that we felt that the 1061 approach i 7 to PRAs was something that should be applied across the i

8 board wherever you did PRAs and that.wasn't evident in the j l 9 first suggestion of the rule on the steam generator, on the 10 proposed changes in the steam generator rule, i

! 11 We suggested that if they are going to use PRA l 12 they ought to be consistent with 1061.

13 CHAIRMAN JACKSON: Well, given that, let me ask 1 i

14 you a question then. So from your understanding, given I i

15 everything you have said of the revised or current approach, 16 would that involve then relaxations in the current tech spec l 17 air criteria?

18 DR. SEALE: Not really. What we really understand i 19 now is that the proposed approach will be to use a generic 20 letter to separate the compliance issues from the voluntary 21 inspection issues or approaches that the utilities might i 22 use, and that if they do any risk assessment that they will 23 base it on the criteria, the approaches set forth in 1061.

l 24 The performance criteria for structures, 25 operational leakage and accident leakage criteria are i

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48 1 essentially _ consistent with what they have now, and ,

2 essentially the structure criteria meet the ASME Code 3 requirements, as we have talked about. [

4 The probabilistic conditional probability for 5 rupture of one or more tubes is a scale going from five 6 times ten to the minus two for one or more tubes to ten to  :

7 the minus three for more than ten tubes.

8 Spontaneous rupture is less than five times ten to 9 the minus two per reactor year.

10 These are criteria that are-set forth in NUREG 11 0844. There is a history of success, if you will, with 12 these criteria, and we think that is probably the 13 appropriate approach to use.

14 CHAIRMAN JACKSON: And so how do you say that t

15 squares with the approach that is being promulgated in the 16 PRAs?

• 17 DR. SEALE: Where they do use, where they come up 18 with alternate approaches based on risk assessment, that 19 that risk assessment should be done in a manner which is 20- consistent with 1061, and in those risk assessments there 21 are proposed performance or levels of allowed frequencies --

22 a thermal challenge frequency, as it is called, for high 23 temperature tube -- for high temperature and elevated 24 differential pressure failures of less than ten to the minus 25 six per reactor year and these approaches then appear to be ,

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49 1 acceptable to us.

2 We haven't gotten the final documents from the -l J

3 Staff, and we will be looking at them in the very near 4 future.

5 It's more of a progress report as to where we are.

6 As I said earlier, went into this with some expectation that 7 a risk approach would be feasible. We' haven't been able to 8 find -- we understand that the Staff hasn't been able to 9 find, to come up with a delivery on that at this point.

10 CHAIRMAN JACKSON: Are there any particular risk -

11 insights that did come out of the Staff's work on the 12 assessment of severe accident induced steam generator tube 13 ruptures that informed -- ,

14 DR, SEALE: Well, I wouldn't call it an. insight, 15 but I would call it a signal as to a concern that we may ,

16 find ourselves addressing more an more often, and that is t 17 that this was one case where what has been an issue that was -

18 strictly in design basis space intruded over into severe 19 accident space in the context of the tube rupture problem as  !

20 a result of a large break LOCA blowdown, and the whole 21 question was exactly what the sequence of events were in the ,

22 load so as to what would fail and in what order and so on. -

23 That brings up a question then as to whether or 24 not in looking at these risk assessments that may be ,

25 appropriate, when you do protrude, if you will, into severe ANN RILEY & ASSOCIATES, LTD.  !

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l 50 l 1 accident space, what is going to be the response to that?

2 Are you going to look at those limited concerns on severe

3. accidents or are you going to rule them out of bounds? '

4 It is a policy issue that we may well have to 5 face. And I think that's the most serious, well let's say a  ;

6 problem that you may very well be concerned with.

7- CHAIRMAN JACKSON: Commissioner Rogers.

8 COMMISSIONER ROGERS: No questions, f

, 9 CHAIRMAN JACKSON: Commissioner McGaffigan.

10 COMMISSIONER McGAFFIGAN: No questions.

11 CHAIRMAN JACKSON: Okay. I think we'll go on.

12 DR. SEALE: Our next speaker is Dr. Powers, and I 13 think you'll find his issues very interesting.

14 . DR. POWERS: I will speak to you a little bit the 15 informed portion of risk informed and performance based 16 regulation. I think you're well aware that when we speak of 17 power operations that the NRC is superbly informed --

18 CHAIRMAN JACKSON: Speak a little more into the 19 microphone.

20 DR. POWERS: And has a tremendous expertise in the 21 risks of power operations. It is, after all, a technology 22 that the NRC developed. It's one that they've nurtured now 23 for two decades. They've honed it with their own analyses, 24 and they've honed it by seeing what the industry can do with ,

i 25 it.

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51 1 There really is no comparable expertise on the 2 risks associated with nuclear power and other modes of 3 operation. Those are the low-power and shutdown modes of l

4 operation. There have been some scoping studies of what .

5 kinds of risks arise during shutdown and low-power 6 operations, and what these scoping studies have shown us is 7 that even when you spread the risks of shutdown operations '

8 over an entire calendar year, you still get results that are l 9 comparable to the risks you have during power operations.

10 What you conclude from that is that the conditional risks of  :

11 shutdown operations must be relatively high compared to the f

12 conditional risks during power operations.

13 What we also know when we look at the records and 14 operational experiences that we have incidents taking place 15 during shutdown and low-power operations. The analyses that.

16 have been prepared.for us for the AEOD show that over 50 '

l 17 percent of all the augmented inspection teams that have been 18 sent to plants by the NRC are to address incidents that have 19 occurred during low-power and shutdown operations. Some of 20 these incidents are relatively serious. We have entered 21 them into the ASP program, and find that they do have very i

22 high conditional core damage probabilities.

23 We're concerned that this situation may actually i

24 get worse, that there are economic pressures on the 25 industry, and they're responding by attempting to shorten ANN RILEY & ASSOCIATES, LTD. l Court Reporters i 1250 I Street, N.W., Suite 300 Washington, D.C. 20005 (202) 842-0034 ,.

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~

I  !

~

j 53

~

l' the periods of shutdown operations. They still have the l l

l 2 same work to do, so they're being asked to do more in ,

3 shorter periods of time, and they may be trying to do it i i

l 4 with fewer people or less-experienced people. At the same i l

5 time, the industry is interested in decreasing the frequency i 6 that it has shut down for refueling and the like. That 7 means there are fewer opportunities to test and exercise  !

these procedures and practices they have during shutdown, 8 l 9 and of course that is the prescription for having an  !

10 increased error rate. We do find that the operators are l

11 under enormous pressures during shutdown operations because l

l 12 there are multiple concurrent evolutions taking place in the l l

l 13 plant. It is a very harassed period of time. ,

I 14 What ACRS has written to you and it has 15 recommended that the NRC needs to develop an understanding I 16 concerning shutdown risks that's comparable to the l

17 understanding that it has during power operations, that the l 18 ACRS understands that this is a very big undertaking, the l

19 technology is not nearly as well developed for analysis of l 20 low-power and shutdown risks, and that the NRC will have to i

21 undertake a development of that technology including a 22 development and understanding of what the success criteria

23 are for shutdown operations.

24 We think you need this understanding.as you embark 25 on this route toward risk-informed regulation. You need l

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53 ,

1 this understanding of risks not because it poses some great 2 benefit to the industry, because what it does is allow you 3 to focus your regulatory actions on those areas that will be 4 truthfully risk-significant.

5 That was essentially the substance of our letter. I 6 I do hope it was clear.

7 CHAIRMAN JACKSON: _Now do you feel that we have an t

8 adequate base experientially, or as-you -- in terms of the  !

9 technology on shutdown risks, PRA's to support ongoing  :

10 rulemaking activities? l t

. 11 DR. POWERS: To support ongoing rulemaking '

i 12 activities on shutdown risk, I don't think you have a risk 13 intuition in this area. I don't think you can-cast your  :

i 14 rules in a quantitative risk framework. We've been making.

l 15 risk-based rules since this agency was formed, but to make  !

i I

16 it quantitative, our arguments have a quantitative i

17 . understanding of the risks during shutdown operations, I ,,

18 don't think you have the technology or the information base  :

19 to do it right now. Even our scoping studies -- they're 20 quite frankly out of date -- the industry understands that 21 this is a problem area for them, and they've instituted l 22 practices that our scoping studies have not reflected. They 23 weren't in place at the time the scoping studies were done, j 24 So if I distinguish between a quantitative 25 understanding of risk and a qualitative understanding of j I

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54 1~ risk, no, you don't have a quantitative understanding of 2 risk to base your decisions on. ,

l

[ 3 CHAIRMAN JACKSON: Is that broad-based, _ or does it 4 relate to, you know, areas of large uncertainty like fire?  ;

, 5 DR. POWERS: In fire or -- you're speaking of fire l

i 6 in general or. fire during the shutdown?

7 CHAIRMAN JACKSON: During the shutdown.

l i 8 DR. POWERS: Fire during shutdown is as big i 9 problem. l 10 CHAIRMAN JACKSON: Right.  !

l 11 DR. POWERS: As you know understanding the risks.  !

l 12 there we quite frankly don't have a good technology for i

13 doing fire in a quantitative risk framework, period. And ,

14 it's no worse nor better in the shutdown operations.

i 15 CHAIRMAN JACKSON: Okay, well I guess what I'm 16 really trying to ask is that relative to shutdown in l 17 particular -- l l

18 DR. POWERS: Um-hum.

l 19 CHAIRMAN JACKSON: Is the effort better focused in 20 areas such as fire risk?

l 21 DR. POWERS: Oh, you're saying you can't do it 22 all, let's do'part of it, and maybe fire is a good place to 23 do part of it?

24 CHAIRMIN JACKSON: I think maybe it is. I'm 25 asking.

i i

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55 4

1 [ Laughter.]

2 DR. POWERS: Is it? I think if you look at the 3 history of incidents, no, the problem is the evolutions in 4 the plants --

5 CHAIRMAN JACKSON: Okay.

6 DR. POWERS: Multiple concurrent activities 7 leading to incorrect valve lineups, incorrect -- conflicting 8 actions where you're having maintenance activities going on 9 in a system that interfaces with a system that you're 10 operating on. I don't think fire is where I would focus my 11 efforts if I had to do a partial job. It's in the multiple 12 concurrent evolutions, and I would pay particular attention 13 to human performance and human error probabilities during 14 these really intense activity times. It's very different 15 than what we're used to in analyzing operator performance 16 under a highly proceduralized single evolutions when the 17 plant is at power.

1. CHAIRMAN JACKSON: Okay.

19 DR. SEALE: You essentially give time because of 20 default trees.

21 CHAIRMAN JACKSON: Yes.

22 DR. POWERS: Yes.

23 DR. SEALE: We don't know what --

24 CHAIRMAN JACKSON: Right. Right.

25 DR. SEALE: I think we know what to do. We just ANN RILEY & ASSOCIATES, LTD.

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56  !

1 need to learn how to do it.

{

2 CHAIRMAN JACKSON: Yes. Commissioner Rogers. I i

l 3 COMMISSIONER ROGERS: Just on this question of the  ;

l i j 4 coupling of low-power and shutdown operations. It seems to  ;

t .

l 5 me that maybe the low-power operations really can be dealt  ;

l 6 with,.I don't know, but within the general framework of

( i i 7 operations, .and any specifics with respect to low power '

8 could be focused on and maybe dealt with more simply.

f 9 Shutdown is, it seems to me, a really different situation.  :

10 You have a lot of different people in the plant, you know,  !

t I

11 it is a very different situation from any kind of power i' 12 operation, f i

l 13 DR. POWERS: I think in any strategy for 6 l

l 14 developing a PRA, for attacking a PRA during the low' power  !

and shutdown operations you would really seriously-think 15 i

l 16 about taking your technology for power operations and 17 evolving it into the lower power operation, I think you l 18 would think about redesigning your technology for shutdown. I i

19 I'm sure that's true.  !

20 I think we have got to take the steps to start

)

21 doing that because this really is occupying an awful lot of '

22 the agency resources and if the benefits that we need to l

23 think about from PRA are not the benefits, the economic 24 benefits to the industry and what-not, it's the focus of our 25 regulations on the places where they have impact, then, my l ANN RILEY & ASSOCIATES, LTD.

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57 1 goodness, here is -- half of our risk is here and we need to 2 focus.

3 CHAIRMAN JACKSON: Commissioner McGaffigan.

4 COMMISSIONER McGAFFIGAN: Do you have any idea of 5 the cost of pursuing a research program to get us to the 6 place you would like us to be, and also the time over which 7 we would need to pursue that program to get to where you 8 want us to be?

9 DR. POWERS: There is probably an integral cost 10 and you probably have a cost-time tradeoff here of some 11 sort.

12 The ACRS tried to be explicit in saying this is 13 not something you can do in a slapdash fashion. You need to 14 take the time to develop your program.

15 I think in our discussions on that, we felt that 16 resolution in this area to the point that you could have 17 something comparable to an analysis of a set of 18 representative plans.

19 You were talking about a period of no less than 20 five years -- some fraction of that in technology 21 development and some fraction of that in the actual conduct 22 of the analyses.

23 We thought it would be a mistake to try to cut 24 corners at this relatively immature level in our 25 understanding, especially of the shutdown aspects of it.

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58 1 I think Commissioner Rogers is absolutely right. '

2 We might be able to evolve into the low power operations 3 with-a few clever analysts but the problem is you have to 4 redefine success criteria for the shutdown sequence, because 5 it is during shutdown you are very likely not to have safety 6 systems. You are very likely to have the containment open 7 to the outside.

8 CHAIRMAN JACKSON: Is it worthwhile to have a 9 focused research program.--

10 DR. POWERS: I think you need one, yes.

11 CHAIRMAN JACKSON: You have to have one?

12 DR. POWERS: I think it is one of your high 13 priority issues right now.

14 COMMISSIONER McGAFFIGAN: And again, how many 15 millions of dollars per year would be -- approximately --

16 DR. POWERS: Well, you know, if you stretch it out 17 to seven years, you probably reduce the million dollar per 18 year by some fraction but it is -- it is not a linear 19 problem.

20 COMMISSIONER McGAFFIGAN: It's on the order of a t'  ;

21 million dollars per year, isn't it?

l 22 DR. POWERS: No , I think it's more on the order of 23 two million dollars -- l l

24 COMMISSIONER McGAFFIGAN: Two million dollars per l 25 year.

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59 1 DR. POWERS: -- is your minimum effort.

2 I think if you go any less than that and you are l

3 just making no progress.  :

4 I think you need --

5 CHAIRMAN JACKSON: What you want to do is you want  ;

/

6 to stoke your research program --

7 DR. POWERS: Right.  ;

8 CHAIRMAN JACKSON: -- then you dollar-load it -- )

Find out your needs first.  !

9 DR. POWERS: Amen.

10 CHAIRMAN JACKSON: Find out your needs so you can 11 stoke it the right way.

12 DR. POWERS: Yes, absolutely. Too often we are 13 designing research programs on what we can do now rather ,

14 than what we ought to be doing.

i 15 CHAIRMAN JACKSON: That's right. l 16 I think we should go on.

17 DR. SEALE: The next speaker here is Dr. Miller on 18 the status of our review of the National Academy report.

19 DR. MILLER: How much time ~do I have? I I

20 CHAIRMAN JACKSON: Three minutes.

21 DR. MILLER: Three minutes, okay, l 22 [ Laughter.)

23 DR. MILLER: I will skip a lot of things then, j i

24 As you know, four years ago this committee f 25 initiated a study by the National Academy of Science to v

f

?

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i 60 1 evaluate the' situation with digital INC. That committee '

2 l

unfortunately didn't start their action until January of '95 3

and gave their first Phase I report in September and there 4

they identified eight issues of importance and significance 5 and which are listed in your briefine b- k and I'll not 6 repeat here.

l 7 The ACRS in October did agree that those issues 8

were amongst the' key issues that would be helpful to digital 9 INC in the future.

10 The Phase 2 report, which again.the charge is 11 listed in your briefing book and I'll skip that, began at i

12 that time and they submitted a written report in January of ,

13 1997 on those issues that they identified previously and ,

14 then we had a presentation and I would say, characterize it 15 as quite valuable dialogue with that committee in March of '

l f 16 '97. .

i 17 During that meeting in March of '97, which is now i 18 just a. couple months ago of course, there were a couple of

'19 other issues that came up which I thought were quite '

20 valuable introduced by individual committee members during 21 the course of that discussion. )

-22 Of course, the Phase 2 report then came up with 39 l 23 - recommendations cn1 those eight issues, of which the Staff i

24 has gone through those recommendations in some detail and I i l

25 have also had the opportunity of going through the Staff's l

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i

I 61 1 disposition of those. t 2 The Staff agreed with 34 of those recommendations  !

3 quite clearly, in fact has even implemented a portion of one i 4 of those into the Standard Review Plan.

5 I'd comment that during this time the Standard ,

6 Review Plan was being updated to incorporate the framework 7 of digital INC -- these things were going on in parallel, ,

8 which was probably a plus or a minus, whichever way you want 9 to look at it.

10 And I have gone through the disposition. Now I 11 have to say one caveat here. The ACRS as a committee has 12 not reached consensus on this report. We have had some 13 debate and so forth and there are certain areas where we are  :

14 going to have to reach consensus in a subcommittee meeting  !

15 in late May.

16 From my point of view I agree with the Staff's 17 disposition on all but one of those recommendations and I 18 had good dialogue yesterday with the Staff, and I think we L 19 are coming to bring closure on even that one.

l i 20 As a consequence, we will have a meeting in May 21 and we are going to address a number of issues including the 22 Standard Review Plan. There will be several issues coming 23 out of this report and I'll just list those, the kind of 24 iccues we are going to be dealing with in May.

25 One of them is Generic Letter 95-02, which ANN RILEY & ASSOCIATES, LTD.

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i l i 62 i

i provides guidance on 5059 for digital upgrades. The second

• 2 is the difference between analog and digital systems, 3 specifically in sampling and also memory-sharing. I 4 The third is a comparison of what the Staff is i 5 doing with a couple of guidelines that were introduced in 6 that report, and that is the FAA guideline and also the ')

7 guideline that has been developed by the Canadians. i 8 The next one is Staff capability. In my view, the 9 Staff Headquarters capability is quite good. I think that  !

1 10 is a substantial change over the last several years at 11 least, in that they are quite good today. There is a plan  !

12 to expand the capability into the regions and I think we

{

13 need to review and make certain the plan is being executed 14 in a reasonably timely fashion.

15 The last issue is one that probably has provided i

16 the most dialogue amongst the ACRS at least, and that's the i

17 balance between the guidance provided for the development l 1

18 process of software versus the final product testing or 1 J

19 product evaluation. In order to facilitate this committee '

20 reaching some consensus on that, during the meeting in May.

21 the Staff has promised me that they will provide a very good {

22 tutorial through example on how they would implement the 23 Standard Review Plan along with the guidance provided by 24 this National Academy report to look at the balance between 25 those, the process of software development and the final i

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63 product evaluation -- so that will be an interesting meeting 1

2 in late May and it promises a lot of interesting discussion 3 amongst this committee in trying to reach consensus on that 4 issue.

\

t 5 To summarize, and I think we have consensus on I f

6 this following statement, the impact of this study, we don't l l

7 believe the findings of the Phase 2 report will lead to any 8 substantial change in the regulatory framework which is 9 being codified in the Standard Review Plan update for l l

10 digital INC.

11 This framework does speak to the major areas of l 12 charge for the Phase 2 report, and that's the areas of 13 criteria for acceptance of digital INC and also the . dance 14 of regulating advanced technology such as digital INC, so 15 that was the charge of the committee but in the sense of the 16 framework being developed it addresses that charge.

17 I would say in my view that some time in June or 18 thereabouts we will have a framework which will put us in a 19 position where the regulatory framework and I think the 20 Staff is moving towards they will have the capability of 21 implementing that framework. It will not inhibit the use of 22 advanced technology in INC systems in nuclear power plants 23 in the next several years, and I am looking forward to 24 seeing a lot of INC with advanced technology going into the 25 power plants over the next several years.

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l 64 1 CHAIRMAN JACKSON: Let me just take -- {

k 2 DR. MILLER: That's all I have but that's a lot of 3- information.

j 4 CHAIRMAN JACKSON: The National Academy's report  !

l l 5 concluded that there are no generally accepted evaluation l .

' 6 criteria for safety-related software.  !

7 Question -- are you saying you agree or disagree, 1

i i 8 and if the agree, then the question is on what basis are  !

l L 9 guidelines and standards set?

t 10 DR. MILLER: Repeat that first part --

l I 11 CHAIRMAN JACKSON: On page 76, Conclusion 1 of the l

12 report --

l 13 DR. MILLER: Right.

14 CHAIRMAN JACKSON: -- the National Academy report, 15 the conclusion was that there are no generally accepted 16 ' evaluation criteria for safety-related software.

17 Do you agree? Does the committee agree or 18 disagree with that conclusion?  !

l 19 DR. MILLER: I would say we agree, yes.

20

CHAIRMAN JACKSON: Okay, so if you agree, then 21 what is the basis for the development of our guidelines, our 22 own guidelines and standards?

23 DR. MILLER: When they stated that, I think they 24 stated it in the sense that they could not guarantee the-25 software would be reliable, but I think in the context of

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65

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1 the software within a total system, I-think that criteria 2: is available -- the total system meaning hardware and 3 software together.

4: I think the National. Academy study and I think 5 pretty much this committee would agree that as long as you 6- look at software in the context of hardware-and it's the

'7 total system, I think the criteria is there.

l 8

Now other members of the committee may want to.  ;

9 speak to that issue.  :

i 10 CHAIRMAN JACKSON: Yes. Dr. Apostolakis?

11 DR. APOSTOLAKIS: I think the fundamental basis ,

t 12 for the development of the guides we have seen or the ,

13 proposed guides is that if you control the process of t

}

14 development of software you will get a very reliable product  ;

i 15 and the Academy does not seem to think that this alone will

'f I

16 do that 17 I think that is the heart of the issue here. As 18 Dr. Miller said, this is something we are still discussing 19 among ourselves and the Staff will come towards the end of ,

i 20 May to educate us a little more about this, but this is the {

' 21 - fundamental thing.

22 It is done in other industries but I think there 23 is a fundamental difference. In other places where they f 24 control the process very well, they have a better 25 understanding of the failure modes of the product, so they [

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66 i 1 know'what to control. '

f 2 CHAIRMAN JACKSON: Are you telling me that the 3 committee has not come down with a position on whether l

1 4

controlling the process of software development gives you 5- the reliability you desire?

6 DR. APOSTOLAKIS: That is correct. I 7 DR. MILLER: Yes.

8 CHAIRMAN JACKSON: And is still under 9 consideration? l 10 DR. APOSTOLAKIS: Yes.

11 DR. MILLER: It is my view, and I say I am not 12 certain where the committee is yet on this, that the Staff 13 has provided the guidance necessary for reviewers to look at 14 the final product and review that final product and its '

l 15 testing of that final product.

I 16 CHAIRMAN JACKSON: But that guidance is referenced

17. to controlling -- the control of the process for the l 18 development of the software.

! 19 DR. MILLER: Also, the guidance has product 20 evaluation.

21 DR. APOSTOLAKIS: Testing.

22 DR. MILLER: More particularly, testing of the l 23 product, and that is spelled out in a Branch Technical i

24 Position, which happened to be Number 14..

25 CHAIRMAN JACKSON: 3ht, but the surrogate test i

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67 1 for reliability is the test of the control of the process 2 for software development?

t 3 DR. APOSTOLAKIS: That is the primary emphasis, I 4 believe, right now, yes, but there is also test of the 5 computer program itself. It's part of the process.

6 DR. KRESS: I think that you are absolutely right 7 though. There is no way to take a piece of software and say 8 how reliable is this piece of software in doing the job that 9 I am asking you to. You cannot do that, and --

10 DR. MILLER: But with regard to the total 11 system --

12 DR. KRESS: You can't really do it in the total 13 system because the software is a part of it and you have to 14 add that part into it, so you cannot do it.

15 The technology does not exist and I think there is 16 no recourse other than to rely on what process --

17 controlling the process. You have to do it and you are 18 doing it on faith.

19 There is no way after the fact to say this process 20 results in a reliable -- you have intuition on it, you have 21 judgment, but it is faith and that's where we're at.

22 CHAIRMAN JACKSON: Is that consistent with the way 23 the FAA does it?

24 DR. KRESS: I'm sorry, I don't know how the FAA 25 does it.

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68 i

1 DR. MILLER: We are going to get an evaluation of 2 the FAA guideline but I believe they follow the same 3 approach generally where you have high control of the 4 process.

5 I want to say that quality control of a process is 6 not different than what we have done in the other areas.

• 7 We have high quality control of the process of l

l 8 development --

9 CRAIRMAN JACKSON: I just want to make sure I 10 understand where we are here.  ;

11 COMMISSIONER ROGERS: I don't have any questions. ,

12 COMMISSIONER McGAFFIGAN: One of the 13 recommendations that the Staff rejected consistent with 14 their interpretation of 5059 was that we loosen up on what 15 an unreviewed safety question is and allow the small changes  ;

t ,

l 16 in risk. '

l 17 Are you in agreement.with the Staff's rejection of  :

18 that recommendation? You said you were largely in agreement 19 at this point.

20 DR. MILLER: I don't want to use the word i 21 " disagreement" -- I believe that further clarification of l

22 the generic letter in the area of system level definition 23 can be done to facilitate our use of digital INC. I think 24 we can do that.  ;

I l 12 5 I had a good discussion with one Staff member 1

i i

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l 69 1 1 earlier this week on that issue.

2 The other issue is defining the difference between l

j 3 simple and complex digital upgrades. I 1

4 I believe the Staff is going to do that in the 5 form of developing some guidelines on use of PLCs.

6 COMMISSIONER McGAFFIGAN: My only comment to you, 7 and this might be future work, but if we get into rulemaking l 8 and so on in 5059, which we may well, following whatever i 9 comments we get on the Staff paper, this might be an area 10 where you all may want to look at the interaction between 11 the change and the rulemaking and the changes proposed in 12 the rulemaking and what you want to accomplish in digital i

13 instrument control.

14 DR. MILLER: Well, I have -- how do I put that?

15 Of course, the ACRS spoke out on that' issue l

16 already but I have some concern about what I saw in that 17 potential rulemaking relating to digital INC.

18 We will definitely be following that issue, as you 19 probably could expect. Does that --

20 CHAIRMAN JACKSON: Okay. I think we should move 21 along.

22 DR. KRESS: Okay. The next on our agenda is Dr.

23 Apostolakis, so George, I'm just going to ask you to hold it 24 to three minutes.

25 CHAIRMAN JACKSON: Three minutes.

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1 DR. APOSTOLAKIS: Okay. Actually, there is no 2 issue here now because we said we didn't like that plan and 3 the Staff relied that we don't like it either, so I 4 understand they.are working on it now and they will come 5 back to us maybe towards the end of June, early July.

6 CHAIRMAN JACKSON: Are you going to be reviewing 7 it prior to June or are you going to wait till the end to i

8 take a look?

9 DR. APOSTOLAKIS

We plan to let the Staff know  !

10 that we would like to do this the way we did the Regulatory ,

i 11 Guides that were just released and preliminary reaction from i g 12 the staff is positive that they would like to come back to l

l 13 us. >

l 14 CHAIRMAN JACKSON: How did you do it relative to 15 the guides that were just released?

l l

l 16 DR. APOSTOLAKIS: Oh, we had very frequent -- -

17 CHAIRMAN JACKSON: In process? i 18 DR. APOSTOLAKIS: In process, right, because they i

19 people are not defensive, you know, it is easier to argue, ,

l 20 so I hope we are going to do this here too, so --

{

l 21 CHAIRMAN JACKSON: I have some questions.

22 DR. APOSTOLAKIS: Okay.

23 CHAIRMAN JACKSON: Have their been any lessons 24 learned from the human reliability modelling performed as 25 part of the IPEs? ,

i I I

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! 71 1 DR. APOSTOLAKIS: IPE? Yes. The state of-the-2 art is in mess. ,

3 I think that was the main message.  !

l l 4' CHAIRMAN JACKSON: All right. What kind of  !

5 database does the Agency have for human errors?

l i

I 6 DR. APOSTOLAKIS: Oh, we have a lot of '

7 incidents --

8 CRAIRMAN JACKSON: Is it a usable database in 9 terms of modelling within this kind of framework?

10 DR. APOSTOLAKIS: I believe the models that are 11 being developed now, yes, they draw on that database.

12 CHAIRMAN JACKSON: And how well is the human 13 performance work coordinated across as well as within ,

l l l 14 offices? l l

15 DR. APOSTOLAKIS: Do I know that?

16 (Laughter.)

17 CHAIRMAN JACKSON: That's an opinion obviously by 18 their facial expressions.

I 19 DR. SEALE: That's part of the problem with the 20 plant.

21 DR. APOSTOLAKIS: The plant had a major problem 22 with that. There was no coordination. Now I don't know 23 whether the research, the ongoing research projects'have 24 that problem too, especially Athena -- I have no idea.

25 CHAIRMAN JACKSON: Maybe that can be spoken of in ANN RILEY & ASSOCIATES, LTD.

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72 1 the context of next week's meeting by the Staff. l

! 2 Okay.

3 DR. SEALE: Okay. We have one last presentation 1

4 and then I'll have a couple comments at the end. '

5 Dana, would you like to mention our letter to 6' Congress?

7 DR. POWERS: Well, let me just be very brief and 8 say that we are by statute required to report to Congress on 9 the state of reactor safety research. We have taken that 10 task very seriously lately because we think the state of 11 research is declining.

12 There is a perception the industry has become l l

13 static and all the problems are solved. We, on the other 14 hand, see an industry that is about to go through big f f

15 changes.  !

i 16 The NRC needs a research program. It served it  ;

3 17 well in the past and will serve it in the future to respond l 18 to those changes so that the NRC is not the bottleneck to j 19 the evolution of the nuclear industry.

20 That was essentially the thrust of our letter to 21 Congress. We will be writing letter of a similar nature  ;

22 each year.

23 We will try to coordinate with you on those, on i 24 producing those letters as best we can. j 25 CHAIRMAN JACKSON: Relative to what you just said, i

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73 1 has the committee reviewed the Staff's proposed criteria for l 2 judging core capabilities?

l 3 DR. POWERS: Certainly I have looked at them. We 4 will in fact be reviewing them in a committee this afternoon l

5 with my presentations on that subject, and I think it is l 6 safe to say that we will have a vigorous discussion on 7 those.

8 CHAIRMAN JACKSON: I see. All right. Oh, I'm 9 sorry, any questions? Commissioner McGaffigan?

! 10 DR. SEALE: Well, I want to thank you very much 11 for your time and your patience with the interest in our l 12 discussions and so on, i 13 I guess one last comment I would make of a l

14 substantial nature is that we try very hard to focus our l

l 15 interests or our questions on PRA, our treatment of PRA, on 16 the benefits that will accrue to the NRC in its attention j l 17 and expenditure of resources necessary to achieve the goals 18 in the safety arena. We feel that the industry's benefits 19 are the interest of these problems. Perhaps that's one 20 reason we suggest that if we want to get some measure of l

21 industry benefit or the possible benefit to industry from 22 PRA applications we should ask them to come up with those 23 definitions, but we try not to get into that particular i

24 arena if we can. We think our emphasis is more appropriate 25 on the NRC, how it does things.

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74

I l 1 We again would like to thank you for your 2 attention and your time, and if you have any questions, 3 elaborating on any of the comments that have been made  ;

4 today, if you'll let our staff know, we'll try to get back  !

l 5 to you.

t 6 CHAIRMAN JACKSON: Let me make a few comments.  ;

l l 7 Let me first thank you for another very informative  ;

8 briefing. You know, we focused on a number of issues that 9 are related to regulatory effectiveness,. and you can see l l

10 that they're linked with our discussions on 5059 and related  ;

11 topics. So I'd encourage you, you know, as you continue to  ;

l l 12 provide us with your perspective, that you be forward-l 13 looking in, you know, bringing developing concerns to the {

! l l 14 Commission's attention in order to help us be prepared for i 15 any future challenges.

i 16 In that light I was particularly interested in the ,

I I' 17 Committee's independent work on acceptance criteria for 18 plant-specific application of safety goals, and deriving  ;

l

19 these lower-tier acceptance criteria, you know, is important i l  ;

l 20 from the point of view of consistency and traceability, and l

21 I hope you continue to pursue these and related activities .

22 in the future.

23 I would also encourage you, to come back to a 24 favorite topic, to take a close look at the adequacy of the 25 guidance being provided by the staff relative to the use of l l

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75 1 uncertainty.

2 DR. SEALE: Good.

3 CHAIRMAN JACKSON: Versus point values in the l

4 decision-making process. You've heard comments from a 5 number of us here.

6 DR. SEALE: Good.

7 CRAIRMAN JACKSON: These are issues to which the 8 Commission and the staff continue to devote considerable 9 time, and I think your involvement would be very helpful.

10 Then finally, in closing, we expect to hear from 11 the staff on the status of the various industry pilots with 12 respect to the topics in question, graded QA, in-service 13 inspection, service testing, and technical specifications, 1

14 at next week's PRA implementation plan briefing.

15 So unless my colleagues have any comments, we're 16 adjourned.

17 We'll take a break. We have another meeting that 18 immediately follows. The break is 2 minutes.

l 19 [Whereupon, at 10:49 a.m., the hearing was 20 concluded.]

21 22 23 i

l 24 i

j 25 ANN RILEY & ASSOCIATES, LTD.

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• CERTIFICATE l

l t

l 1 l 1 1

! i This.is to certify that the attached description of a meeting i of.the U.S. Nuclear Regulatory Commission entitled:

i ,

l TITLE OF MEETING: MEETING WITH ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS) - PUBLIC .

l  :

MEETING I PLACE OF MEETING: Rockville, Maryland i

1 l DATE OF MEETING: Friday,-May 2, 1997

(

l was held as herein appears, is a true and accurate record of the meeting, and that this is the original transcript thereof l taken stenographically by me, thereafter reduced to l

typewriting by me or under the direction of the court reporting company l

Transcriber: C hu 6te n tu i fe,-t4 n CT

! Mark Mahoney Reporter:

1 i

i j

i 1 . . .

e ven 8 'o UNITED STATES g

8 o NUCLEAR REGULATORY COMMISSION

$ ,U ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ WASHINGTON, D. C. 20555 o

l  %,*****s*g April 22, 1997 l

i . MEMORANDUM TO: John C. Hoyle l Secretary of the Commission

^

FROM: John T. Larkind Txe/ ffvtive Director Advisory Committee on Reactor Safeguards h

SUBJECT:

ACRS MEETING WITH THE NRC COMMISSIONERS, MAY 2, 1997-SCHEDULE / BACKGROUND INFORMATION t

The ACRS is scheduled to meet with the NRC CommissionI9s between 9:00 and 10:30 a.m. on Friday, May 2, 1997, to discuss the items listed below. Background materials related to these items are l attached.

A. Introduction - NRC Chairman 9:00 - 9:05 a ca.

i B.1 (a) Risk-Informed, Performance-Based 9:05 - 9:20 a.m.

Regulation and Related Matters Dr. Apostolakis (pp. 1-9)

(b) Risk-Based Regulatory Acceptance 9:20 - 9:35 a.m.

Criteria for Plant-Specific Application of Safety Goals Dr. Kress (; p p. 10-37)

2. Proposed Regulatory Approach 9:35 - 9:45 a.m.

Associated with Steam Generator .

Integrity Dr. Seale (; p p. 38-50) l 3. Low-Power and Shutdown 9:45 - 10:00 a.m.

l Operations Risk Dr. Powers (pp. 51-62)

4. Status of ACRS Review of National 10:00 - 10:10 a.m.

Acadeny of Sciences / National Research Council Phase 2 Study Report on Digital Instrumentation and Control Systems Dr. Miller (pp. 63-68)

5. Human Performance Program Plan 10:10 - 10:20 a.m.

Dr. Apostolakis (pp. 69-77) l l

l

1 s' ,

John C. Hoyle 6. ACRS Report to Congress on 10:20 - 10:25 a.m.

Nuclear Safety Research and i Regulatory Reform Dr. Powers (pp. 78-86)

C. Closing Remarks - NRC Chairman 10:25 - 10:30 a.m.

Attachment:

As stated cc: ACRS Members ACRS Technical Staff l

I 1

I I

I e

i l

i i

1 l

I 1

l I

i

, I l ITEM B.1 (a): l RISK-INFORMED, PERFORMANCE-BASED REGULATION AND RELATED MATTERS (DR. APOSTOLAKIS) l l

l l

t J

., i l

ITEM B.1 M: RISK-INFORMED. PERFORMANCE-BASED REGULATION AND I RELATED MATTERS I During the December 6,1996 meeting with the Commissioners, the ACRS discussed the staff's I approach to codify risk-informed, performance-based regulation and related matters through l development of Standard Review Plan (SRP) sections and associated regulatory guides.

I 1

Subsequent to the December 6,1996 meeting with the Commission, the ACRS has continued l its discussions with the staff regarding the proposed SRP sections and regulatory guides. The l ACRS met with the staff and industry representatives on February 6-8 and March 6-8,1997, to continue its discussion of the proposed SRP sections and regulatory guides. The ACRS Subcommittee on Probabilistic Risk Assessment (PRA) also heard presentations by and held discussions with representatives of the staff and of the industry on January 28, February 20 l and 21,1997. All ACRS members attended the Febmary 20-21 Subcommittee meeting.

During these meetings, the ACRS reviewed the following documents:

• Draft Regulatory Guide DG-1061, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," dated February 28,1997

• Draft Regulatory Guide DG-1062, "An Approach for Plant-Specific, Risk-Informed, Decision Making: Inservice Testing," dated February 25,1997

• Draft Regulatory Guide DG-1064, Revision 4, "An Approach for Plant-Specific, Risk-Informed Decision Making: Graded Quality Assurance," dated February 26,1997

~

Draft Regulatory Guide DG-1065, Revision 5, "An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications," dated Febmary 24,1997

• Draft SRP Chapter 19, Revision L, "Use of Probabilistic Risk Assessment in Plant-Specific, Risk-Informed Decisionmaking: General Guidance," dated March 3,1997

• Draft SRP Section 3.9.7, Revision 2C, " Standard Review Plan for the Review of Risk-Informed Inservice Testing Applications," dated February 25,1997

• Draft SRP Chapter 16.1, Revision 12, " Risk-Informed Decision Making: Technical Specifications," dated February 24,1997 1

l

O i The ACRS was provided with draft NUREG 1602, " Standards for Probabilistic Risk Assessment (PRA) to Support Risk-Informed Decisionmaking." Since it was being revised, the staff requested that the ACRS not comment on this document at this time. The ACRS plans to review this document after receiving a revised document.

The ACRS provided a report to the Commission on March 17, 1997. In this report the Committee made several comments and recommendations, including the following:

• The draft Regulatory Guide DG-1061 (General Guidance) and the associated SRP Chapter 19 that provide guidance for making risk-informed changes to the current licensing basis of individual plants constitute a significant achievement. These documents, and in panicular the stated principles, provide the foundation for risk-informed regulatory philosophy that can better focus resources and can lead to a more coherent regulatory st:ucture.

The staff formulated a list of questions to elicit public comments needed to refine and improve these draft documents. The ACRS agreed that these questions highlight important issues that need to be addressed. The ACRS recommended that the staff issue these documents for public comment.

• The draft Regulatory Guides DG-1062 (Inservice Testing) and DG-1065 (Technical Specifications) and the associated SRP sections have been developed consistent with the principles in DG-1061 (General Guidance). The ACRS recommended that these documents be issued for public comment.

The ACRS viewed proposed guidelines for allowed outage times to be acceptable in that they were based on incremental conditional probability of core damage and large, early release. The ACRS stated that such guidelines could also include limits on the maximum conditional annual core damage frequency (CDF) and large, early release frequency (LERF). This alternative is one of the questions included in the proposed Federal Register notice, and the ACRS believes that final resolution of this issue can be postponed until public comments on this matter have been received.

The ACRS considered DG-1064 (Graded Quality Assurance) to have taken an']

unnecessarily timid approach toward focusing stringent quality assurance activities on highly risk-significant systems, structures, and components (SSCs) and defining adequate, but less exacting, quality assurance demands on areas oflow risk significance.

The Committee stated that there should be a clear justification based on PRA or other compelling reasons for classifying SSCs as belonging to the high-safety-significant category.

1 2 ,

1

I The ACRS stated that issuance of this draft Regulatory Guide, in this form, could reinforce the suspicion that risk-informed regulation is simply an additional layer of 1 regulation imposing burden without tangible benefit. Since the staff agreed to revise this document to address the ACRS concerns and, since there is intense industry interest, the Committee stated that it has no objection to the staff proposal for issuing i the revised document for public comment.

I i

• The ACRS believes that the successful implementation of the risk-informed,  !

parformance-based regulatory philosophy will require a change in culture for both the NRC staff and the industry. The ACRS recommends that a vigorous program be established to communicate the risk-informed philosophy through workshops and other means planned by the staff.

• The ACRS recommends that the Commission consider issuing a statement inviting ,

licensees to propose new and innovative approaches to risk-informed, performance- i based regulation using the concepts articulated in DG-1061 (General Guidance). The

! review and approval processes may need to be revised to provide timely responses to licensee submittals, which will necessarily cross disciplinary and organizational lines.  ;

l l

• DG-1061 (General Guidance), Appendix B, provides a method for estimating LERF m l the absence of a Level 2 PRA. The ACRS recommends that some approaches also be l developed for estimating the contributions of external events to CDF and LERF, as l

l well as from low-power and shutdown operations when detailed PRAs are not ,

available. l The ACRS noted that frequent interactions between the ACRS and the staff on the draft SRP sections and associated regulatory guides were both valuable and constructive, and the staff was very cooperative during this long process. j The ACRS plans to continue its review of proposed SRP sections and regulatory guides for risk-informed, performance-based regulation. The ACRS Subcommittee on PRA plans to meet I with the staff to discuss the proposed SRP section and associated regulatory guide for risk-informed inservice inspection in June 1997 with full Committee review during the July 9-11, 1997 ACRS meeting.

The Committee plans to meet with the staff to discuss the PRA Implementation Plan during I its June 11-13, 1997 meeting, with emphasis on risk-informed initiatives in training and inspection. The ACRS also plans to meet with the staff to discuss ongoing staff initiatives for risk-based analysis of reactor operating experience nd some special studies.

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Attachment:

• Report dated March 17, 1997, from R. L. Scale, ACRS Chairman, to Shirley Ann l Jackson, NRC Chairman,

Subject:

" Proposed Standard Review Plan Sections and Regulatory Guides for Risk-Informed, Performance-Based Regulation" (pp. 5-9) i l

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. [ o NUCLEAR REGULATORY COMMISSION UNITED STATES f e

'f I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS cAsmucTow, c. c. nosss e,,

March 17,1997 The Honorable Shirley .\nn, Jackson Chairman i

U.S. Nuclear Regulatory Commission l

Washington, D.C. 20555-0001

SUBJECT:

PRCJOSED STANDARD REVIEW PLAN SECTIONS AND REGUIATORY GUIDES FOR RISK-INFORMED, PERFORMANCE-BASED REGULATION During the 437th, 438th, and 439th meetings of the Advisory Committee on Resctor Safeguards, December 5-7, 1996, February 6-8, and March 6-8, 1997, respectively, we met with representatives of the NRC staff, industry, and other interested parties to review the proposed Standard Review Plan (SRP) sections, Regulatory Guides, and other matters associated with risk-informed, performance-based regulation. We discussed the staff's approach to codify risk-informed, performance-based regulation into a general guidance SRP section and an associated Regulatory Guide, as well as related documents for technical specifications, inservice testing, and graded quality assurance. We also discussed industry views and these matters. In addition, our initiatives related to .

Subcommittee on Probabilistic Risk Assessment (PRA) met with the staff and industry representatives to discuss these documents on October 31, November 1, 21, 22, 1996, and January 28, February 20 and 21, 1997. We also had the benefit of the documents referenced.

conclusions and Recomunendations

1. The draft Regulatory Guide DG-1061 (General Guidance) and the associated SRP Chapter 19 that provide guidance for making risk-informed changes to the current licensing basis of individual plants constitute a significant achievement. They, provide the and in particular the stated principles, foundation for risk-informed regulatory philosophy that can better focus rerourcesThese and can lead to a more coherent documents should be issued for regulatory strucwre.

public comment. The staff has formulated questions that should elicit the public comments needed to refine and improve these draft documents.

2. The draft Regulatory Guides for application of risk-informed decisionmaking to technical specifications (DG-1065) and in-service testing (DG-1062) have been developed consistent with 5

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i i' 2 i i i the principles articulated in the General Guidance (DG-1061) .  ;

} , These Regulatory Guides and associated SRP sections should be  :

i issued for public comment. The proposed guidelines for "

i acceptability of allowed outago time changes are based on the

• J incremental conditional probakility of core damage and large, I early release. such guidelinos could also include limits on i the maximum conditional am ual core damage frequency (CDF) and

large, early release frequency (LERF). This alternative is i one of the questions included in the proposed Federal Reaister j notice, and we believe that final resolution of this issue can 1 be postponed until public comments on this matter have been i . received.

i j 3. The version of the Regulatory Guide that we reviewed for j application of risk-informed decisionmaking to graded quality i

assurance (DG-1064) took an unnecessarily timid approach

toward focusing stringent quality assurance activities on j highly risk-significant systems, structures, and components

] (SScs) and defining adequate, but less exacting, quality '

i assurance demands on areas of low risk significance. There '

i should be a clear justification based on PRA or other j

compelling reasons for classifying SSCs as belonging to the j high-safety-significant category. Issuance of this draft j Regulatory Guide in this form for public comment could

erroneously reinforce the wide-spread suspicion that risk-

{ informed regulation is simply an additional layer of i regulation imposing burden without tangible benefit. The i' staff is currently working to revise this document, and we are confident that the revised version will, in large measure, address our concerns. Since there is intense industry interest in this Regulatory Guide, we have no objection to the staff's proposal for issuing this document for public comment.

' 4. The successful implementation of the new regulatory philosophy will require a change in culture for both the NRC staff and i the industry. A vigorous program should be established to j communicate the risk-informed philosophy through workshops and ,

i other means planned by the staff. '

l i 5. The Commission should consider issuing a statement inviting j licensees to propose new and innovative approaches to risk-i informed, performance-based regulation using the concepts j ,

articulated in the General Guidance (DG-1061) . The review and approval processes may need to be revised to provide timely responses to licensee submittals, which will necessarily cross

! disciplinary and organizational lines.

j 6. Appendix B of DG-1061 provides a method for estimating LERT in i the absence of a Level 2 PRA. We recommend that some

} approaches also be developed for estimating the contributions l 1 i

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l of external events to CDF and LERF, as well as from low-power and shutdown operations when detailed PEAS are not available.

( Discussion It has been about 22 years since the Reactor Safety Study (WASH-1400) introouced PRA to the reactor safety community. During this  :

time, PRA methods, especially those for internal events during power operations, have natured to the point that PRA insights are increasingly being utilized in risk management both by the staff and licensees. However, formal guidance as to how PRA results can be used in the regulatory arena has been lacking. The documents

! that the staff has prepared provide such guidance. They constitute i l

l a major step forward in the development of a more risk-informed  !

regulatory process.

Formulation of the guidance in terms of a basic set of principles -

creates a foundation for the new regulatory philosophy. We believe this to be a sound and significant achievement. It provides the starting point for the integration of traditional engineering approaches to safety, such as defense-in-depth, and the new probabilistic approach. The implementation of the General Guidance will evolve as experience is gained. We are confident, however, that a good start has been made.

The efforts to understand how the concepts of defense-in-depth and safety margins can be considered in the context of PRA must be applauded and encouraged. They provide very useful insights I

regarding the intent of these cornerstones of traditional reactor safety philosophy and the extent to which they are reflected in the PRA results.

We agree with the use of an " integrated" process in risk-management situations. It is clearly recognized that decisionmaking cannot rely solely on numerical results from either the PRA or more traditional approaches. We note that this integrated approach to decisionmaking is akin to the concept of the inclusion of

" deliberation" in reaching risk-management decisions, as discussed in a recent report by the National Research Council Committee on Risk characterization.

A first reading of the proposed SRP sections and associated Regulatory Guides creates the impression that they impose an onerous burden and are difficult to understand. We believe that the potential benefits clearly outweigh this burden and merit the effort to implement the new philosophy.

At this time, many of the staff and the industry may still not believe that risk-informed regulation is real or may have difficulty in making the transition to risk-informed decision-making. We are, therefore, very pleased to hear that the staff I

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j plans to organize workshops and public meetings to explain the new j regulatory philosophy. ,

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Interactionitswith industry is needed to ensure that the industry realizes .

responsibility to provide the staff with the  !

information needed te make meaningful risk-informed decisions. To i a large extent, the benefits of a risk-informed approach will be in '

j proportion to the attention to accuracy and completeness of the

industry's PRAs. There probably are licensees that have not yet J

done enough with their current PRAs to be able to garner j significant benefit from a risk-informed approach to regulation.

The new approach may, in the beginning, require additional industry i

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. effort.

ciontext of This is understandable and should be considered in the potential <

increases in safety and reductions in l

regulatory burden.

Graded quality assurance is a quintessential subject for j application of risk-informed decisionmaking. Risk information j should be the rational basis for adjudicating the level of quality j

assurance effort needed to provide confidence that SSCs will j perform their safety functions reliably. The staff is currently

planning to use risk information only to reduce quality assurance

] requirements for ssCs in the low-safety-significant category. We j believe there is a better approach to using risk information to j classify SSCs according to quality assurance needs than that described in the draft Regulatory Guide. Greater discrimination among the quality assurance needs will better focus licensee and j

regulatory attention on risk-imporunt topics. Such a focus may not be achieved by simply using risk information to define reductions in licensee burdens associated with quality assurance

{ for low-safety-significant items.

i The approach proposed in Appendix B of DG-1061 for estimating LERF l i in the absence of a Level 2 PRA needs to be supported with {

i additional documentation. Although this approach may be i l appropriate for screening purposes, additional probabilistic i

analyses using plant-specific values may be necessary for plants

that do not meet the LERF guidelines.

4 l We express our appreciation for the staff's cooperation during this long process. We have had excellent discussions of both concepts and methods during our meetings. The staff was always willing to listen and debate with us. The frequent interactions between the ,

i staff and the Committee were very valuable and constructive. l 1

{ sincerely, j R. L. seale

Chairman i

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1. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG- .

1061, "An Approach for Using Probabilistic Risk Assessment in i l

Risk-Informed Decisions on Plant-Specific Changes to the current Licensing Basis," dated February 28, 1997.

2. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-1062, "An Approach for Plant Specific, Risk-Informed, Socision l Making: Inservice Testing," dated February 25, 1997.  ;

j 3. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG- l l 1064, Revision 4, "An Approach for Plant-Specific, Risk-i Informed Decision Making: Graded Quality Assurance," dated i

February 26, 1997.

4. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-  !

! 1065, Revision 5, "An Approach for Plant-Specific, Risk-l Informed Decision Making: Technical Specifications," dated  ;

l February 24, 1997.

l

5. U.S. Nuclear Regulatory Commission, Draft Standard Reviev Plan, Chapter 19, Revision L, "Use of Probabilistic Risk l

Assessment in Plant-Specific, Risk-Informed Decisionmaking:

General Guidance," dated March 3, 1997.

l 6. U.S. Nuclear Regulatory Commission, Draft Standard Review

! Plan, Chapter 3.9.7, Revision 2C, " Standard Review Plan for {

l the Review of Risk-Informed Inservice Testing Applications,"

dated February 25, 1997.

7. U.S. Nuclear Regulatory Commission, Draft Standard Review Plan, Chapter 16.1, Revision ' 12, " Risk-Informed Decision Making: Technical Specifications," dated February 24, 1997.

8. U. S. Nuclear Regulatory Commission, NUREG-75/014, " Reactor l Safety Study, An Assessment of Accident Risks in U.S. Nuclear i

Power Plants," WASH-1400, October 1975.

9. National Research Council report, " Understanding Risk, Informing Decisions in a Democratic Society," 1996.

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ITEM B.1 (b):

RISK-BASED REGULATORY ACCEPTANCE CRITERIA FOR PLANT-SPECIFIC APPLICATION OF SAFETY GOALS (DR. KRESS) ,

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ITEM B.1 M: EISK-BASED REGULATORY ACCEPTANCE CRITERIA FOR PLANT-SPECIFIC APPLICATION OF SAFETY GOALS The Committee provided a repon to the Commission, dated November 18,1996, regarding )

" Plant-Specific Application of Safety Goals." In this report the Committee made several  !

comments and recommendations, including the followmg: l 1

• The safety goals and subsidiary objectives can and should be used to derive l guidelines for plant-specific applications. It is, however, impractical to rely l exclusively on the Quantitative Health Objectives (QHOs) for routine use on I an individual plant basis. Criteria based on core damage frequency (CDF) and large, early release frequency (LERF) focus more sharply on safety issues and i can provide assurance that the QHOs are met.

• The Safety Goals quantified "how safe is safe enough" for the population of i U.S. plants. For an individual plant, however, the acceptable level of risk is l determined by the concept of " adequate protection," which in the final analysis l means compliance with the body of regulations. Risk-informed analyses would provide a more rational basis for making regulatory decisions regarding plant-  !

specific requests for enmptions from the rules or for changes to the licensing I basis, and the acceptability of new regulations.

• In the longer term, the agency should niove beyond the evaluation of risk associated with proposed changes to individual plant licenses and apply the  !

Safety Goals to assess the acceptability of plant-specific risk. This could be done in terms of the QHOs, along with the CDF, or in terms of the CDF and LERF. To use the QHOs directly, it would be necessary to have full-scope Level 3 probabilistic risk assessments (PRAs).

The Committee met with the Commissioners on December 6,1996, and discussed the report mentioned above. During that meeting, the Committee committed to provide an example of how risk-acceptance criteria could be developed directly form the Safety Goals. In addition, in a Staff Requirements Memorandum datedJanuary 14,1997, the Commission asked for the ACRS views on the relationship between the concept of " adequate protection" as used in NRC regulations, and the NRC Safety Goals, from the standpoint of levels of risk.

The Committee discussed the concept of " adequate protection" and other related matters and provided a repon to the Commission on April 11, 1997 regarding " Risk-Based Regulatory Acceptance Criteria for Plant-Specific Application of Safety Goals." In this report, the following two main points were emphasized.

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l l 1. The lower tier risk-acceptance criteria (CDF and LERF), now being proposed in draft

Regulatory Guide DG-1061, "An Approach for Using Probabilistic Risk Assessment

! in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," j for use in making decisions regarding requested changes to a licensee's current licensing basis, should be derived directly from the prompt fatality QHO and should be of such ,

a value as to bound all current sites.  !

2. In the long run for enforcement purposes, the prompt fatality QHO should be )

considered as the quantification of a risk level to replace " adequate protection."

l Some of the discussion points concerning the derivation of lower tier risk-acceptance criteria )

included: i l

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• A risk-informed, performance-based regulatory system ought not be implemented l without the existence of top-level risk acceptance criteria. The obvious choices for ,

these criteria are the NRC Safety Goals.

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• The subsidiary CDF goal should be elevated to the status of a fundamental goal.  ;

l Elevating the CDF subsidiary goal to the status of a fundamental goal can be i j considered as a defense-in-depth principle that provides balance between prevention and  !

l mitigation.

[

• The understanding of risk associated with low-power and shutdown operations, or l accidents initiated by external events in which emergency response is impeded, is not l yet sufficient to draw definitive conclusions concerning the limiting QHO in these  !

l situations.  :

Some of the discussion points concerning adequate protection include:

• Since each licensed plant must, by definition, provide adequate protection, the licensed l

l plant that poses the highest level of risk places a bound on the quantified level of risk l to be associated with " adequate protection."

• Within the spectrum of risk, it is likely that there are plants with risk levels above the ,

L Safety Goals and other plants with risk levels below. If this is indeed the case, a single >

risk level that bounds " adequate protection" would be a risk level greater than the Safety Goal level.  !

• A long-term objective of replacing the " adequate protection" concept with a well articulated and quantified " acceptable level of risk" if achievable, would enhance the public's understanding and acceptance of the regulatory process and would lead to a -

more uniform level of protection for all individuals living in the vicinity of nuclear ,

plants.

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1 l Additional comments attached to the April 11, 1997 ACRS report, provide examples of approaches that could be used to quantify lower tier acceptance criteria (i.e., LERF, or CDF  :

l and conditional containment failure probability), which would ensure that the early fatality +

l QHO is met at each site.

Attachments-i e Repc c dated November 18,1996, from T. S. Kress, ACRS Chairman, to Shirley Ann Jackson, NRC Chairman,

Subject:

Plant-Specific Application of Safety Goals (pp.13-15) l

• Staff Requirements Memorandum, dated January 14,1997, from John C. Hoyle, Office

, of SECY, to John T. Larkins, ACRS Executive Director (pp.16-17) ,

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• Report dated April 11,1997, from Robert L. Scale, ACRS Chairman, to Shirley Ann Jackson, NRC Chairman,

Subject:

Risk-Based Regulatory Acceptance Criteria for l Plant-Specific Application of Safety Goals (pp.18-37) l l

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mg'o UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

.I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

. WASHINGTON, D. C. 20655

. \ * * * * * /,

November 18, 1996 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Dear Chairman Jackson:

j

SUBJECT:

PLANT-SPECIFIC APPLICATION OF SAFETY GOALS During the 436th meeting of the Advisory Committee on Reactor Safeguards, November 7-9, 1996, we discussed the application of Safety Goals on a plant-specific basis. This subject was also discussed at meetings of our Joint Subcommittees on Probabilistic Risk Assessment and Plant Operations on July 17-18, 1996, and of our Subcommittee on Probabilistic Risk Assessment on August 7, 1996. We also had the benefit of the documents referenced.

In a Staff Requirements Namorandum dated June 11, 1996, we were requested to provide recommendations on how the Commission's Safety  ;

Goals and safety Goal Policy should be revised to make them  !

acceptable for use on a plant-specific basis, j l

The Safety Goal Policy Statement made it clear that the i Quantitative Health Objectives (QHos) and the subsidiary Core l Damage Frequency (CDF) goal were to provide standards for the NRC staff to judge the overall effectiveness of the regulatory system.

That is, if the risk posed by the population of plants on the average proved to be less than the Safety Goals, then the staff (and presumably the public) would deem that the regulatory system had functioned appropriately to protect the health and safety of the public.

The Safety Goals quantified "how safe is safe enough" for the population of U. S. plants. For an individual plant, however, the acceptable level of risk is determined by the concept of " adequate protection," which in the final analysis means compliance with the body of regulations. Risk-informed analyses would provide a more rational basis for making regulatory decisions regarding plant-specific requests for exemptions from the rules or for changes to the licensing basis, and the acceptability of new regulations.

In our August 15, 1996 report, we stated: "the safety goals and subsidiary objectives can and should be used to derive guidelines for plant-specific. applications. It is, however,. impractical to rely exclusively on the Quantitative Health Objectives (QHos) for

' routine use on an individual plant basis. Criteria based on core 13

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damage frequency (CDF) and large, early release frequency (LERF) focus more sharply on safety issues and can provide assurance that j the QHos are met."

In developing plant-specific criteria, it is important to consider the regulatory needs in the near future and to ensure that the i

process will be evolutionary rather tlan so revolutionary that it l migt.t discourage the licensees from using this approach. It j appears that nost of the anticipated licensee requests for changes i to thei.- current licensing basis will deal with Level 1 1 probabilistic risk assessment (PRA) issues, e.g., inservice j inspection, extension of allowed outage times. Furthermore, most

) licensees have only recently familiarized themselves with Level 1 j PRA methodology for the narrow regime of power operations. They j are just beginning to integrate findings of such Level 1 risk assessments with the safe operation of their plants. Even the NRC 1 staff is still coming to grips with the implications of Level 1 risk assessment results for regulation of nuclear plants. Many

! licensees do not have access to the technologies for facile conduct j of full-scope Level 2 or Level 3 PRAs that treat power operations, i low power / shutdown operations, as well as accidents initiated by i external events. Commonly accepted standards for such extensive, j in-depth analyses do not exist.

An evolutionary and pragmatic approach for using Safety Goals on a 4 plant-specific basis would be to use the CDF as the primary

{ criterion for evaluating proposed changes along with a qualitative

! or quantitative evaluation of the possible Level 2 and Level 3 PRA issues raised by these changes. For a quantitative analysis, the following two options are offered:

) 1) Full-scope Level 2 PRA (with. fission product transport 1 capability).

To use this option, a conservative value for a LERF criterion must  !

be determined. This value, along with the CDF criterion, will provide an acceptable basis for decisionmaking. We note that both

< the NRC staff and the Electric Power Research Institute, in its, "PSA Application Guide," are proposing the use of LERF as an acceptance criterion.

. 2) Full-scope Level 2 PRA (without fission product transport j capability).  :

! To use this option, conservative values for early containment i failure frequency criteria for different reactor designs must be i determined. These values, along with the CDF criterion, will j provide an acceptable basis for decisionmaking.

l In the longer term, we believe the agency should move beyond the j evaluation of risk associated with proposed changes to individual j plant licenses and apply the Safety Goals to assess the j 14 i

3

acceptability of plant-specific risk. This could be done in terms l of the QHos, along with the CDF, or in terms of the CDF and LERF.

l To use the QHos directly, it would be necessary to have full-scope

! Level 3 PRAs. We believe that the use of Level 3 PRAs in the i future should be encouraged.

Sincerely, l

.J 5 /w T. S. Kress Chairman l

References:

1

1. Staff Requirements Memorandum dated June 11, 1996, from John Hoyle, Secretary, NRC, to John T. Larkins, Executive Director, l ACRS,

Subject:

Meeting with ACRS, Friday, May 24, 1996 l 2. ACRS report dated August 15, 1996, from T. S. Kress, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,

Subject:

Risk-Informed, Performance-Based Regulation and Related Matters l 3. Electric Power Research Institute Report TR-105396, "PSA Application Guide," prepared by ERIN Engineering and Research, Inc., August 1995 i

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s UNITED STATES l f , NUCLEAR REGULATORY COMMISSION l WASHINGTON, D.C. 20555-0001 IN RESPONSE, PLEASE On*** January 14, 1997 REFER TO: M961206A l CFFICE OF THE SECRETARY l

l MEMORANTUM TO: John T. Larkins, Executive Director Advi o Co. ittee on Reactor Safeguards FROM: Jo . oy , Secretary

SUBJECT:

STAFF REQUIREMENTS - MEETING WITH ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS), 9:30 A.M., FRIDAY, DECEMBER 6, 1996, COMMISSIONERS' CONFERENCE ROOM, ONE WHITE FLINT NORTH, ROCKVILLE, MARYLAND (OPEN TO

. PUBLIC ATTENDANCE) l The Commission was briefed by the ACRS on the following topics:

1. Digital Instrumentation and Control Systems

2. Office of Nuclear Regulatory Research (RES) Plan for Upgrading Thermal-Hydraulic Codes l

3. Risk-Informed, Performance-Based Regulation and Related l Matters l

4. Potential Use of IPE/IPEEE Results to Compare the Risk i of the Current Population of Plans with the Safety l Goals

5. Use of Safety Goals on a Plant-Specific Basis

6. Use of RuleNet in the Regulatory Process.

I ACRS should continue to be forward-looking to bring developing l concerns to the Commission's attention and continue follow up on l issues such as digital IEC and use of Safety Goals for regulatory I

purposes. In this regard, the commission would be interested in the ACRS views on the relationship between the concept of

" adequate protection," as used in the NRC regulations, and the NRC safety goals, from the standpoint of levels of risk.

On the issue of I&C design process and acceptable product performance, the staff was asked to reduce the use of standards i referenced and to consider how a process and product specific QA could track requirements and ensure the acceptability criteria is pertinent and sufficient.

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l cc: Chairman Jackson Commissioner Rogers Commissioner Dieus l Commissioner Diaz l Commissioner McGaffigan 1 OGC l OG 01G Office Directors, Regiuns. ACRS, ACNW, ASLBP (via E-Mail)

PDR - Advance l DCS - P1-24 ,

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/' 'o UNITED STATES g

! o NUCLEAR REGULATORY COMMISSION

$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS a WASHINGTON. D. C. 20555 April 11, 1997 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

RISK-BASED REGULATORY ACCEPTANCE CRITERIA FOR PLANT-SPECIFIC APPLICATION OF SAFETY GOALS l In our December 6,1996 meeting with the Commission, we committed to provide an example of how risk-acceptance criteria could be developed directly from the Safety Goals. Additionally, in a Staff Requirements Memorandum dated January 14, ,

l 1997, the Commission asked for our views on the relationship between the concept  !

of " adequate protection," as used in the NRC regulations, and the NRC Safety Goals, from the standpoint of level of risk. 1 During the 440th meeting of the ACRS, April 3-4, 1997, we completed oar deliberations on plant-specific application of NRC Safety Goals and the relationship between the concept of " adequate protection" and the Safety Goals.

In our November 18, 1996 report on this subject, we stated that "the safety goals and subsidiary objectives can and should be used to derive guidelines for plant-specific applications." We noted that full-scope Level 3 probabilistic risk assessments (PRAs) would be necessary to use the quantitative health objectives (QH0s) directly to assess the acceptability of plant-specific risk. We also stated that this assessment of risk could be done in terms of the QH0s, along with the core damage frequency (CDF), or in terms of the CDF and large, early release frequency (LERF).

This report further discusses the need for plant-specific application of risk-acceptance criteria and the appropriateness of these criteria being derived from i the Safety Goal QHO on early fatalities. The additional comments to this report provide examples of approaches that could be used to quantify lower tier acceptance criteria (i.e., LERF, or CDF and conditional containment failure probability) that will ensure that the early fatality QHO is met at each site.

l Quantification of the LERF at each site is needed to ensure the appropriateness l

of the choice of the LERF acceptance criterion proposed in draft Regulatory Guide DG-1061 and draft Standard Review Plan sections that support risk-informed, performance-based regulation.

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2 Need for Plant-Soetific Aeolication The Safety Goal Policy Statement makes it clear that the QH0s and the subsidiary goal on CDF were intended only to provide standards for the NRC to judge the overall effectiveness of its regulatory system. The Policy Statement 1 specifically preclude:, enforcement of the Safety Goals on a plant-specific basis.  !

l In the development of draft Regulatory Guide DG-1061 and the associated draft l Standard Review Plan sections in support of risk-informed, performance-based i l regulation, the staff has found it necessary to propose risk-acceptance <

l guidelines that can be applied on a plant-specific basis. These guidelines would be used, along with other considerations and inputs, for making judgments on the l acceptability of requested changes to a licensee's current licensing basis.

Reviewing plant-specific license amendments by using risk-acceptance guidelines is a positive action toward risk-informed, performance-based regulation.

l  !

We also note that, in the longer term, the Commission may want to consider having

a quantified acceptable risk level to replace the current concept of " adequate protection." This risk level could eventually serve as an objective risk-acceptance criterion for many enforcement decision

;.

l Risk-informed. Performance-Based Reculation The Commission has directed the staff to increase the use of PRA in the  !

regulatory process. We have endorsed this because we believe that a risk-  ;

informed, performance-based regulatory approach will lead to increased coherence I in the regulatory system, to enhanced decision-making ability, and to technically defensible bases for granting regulatory relief.

A risk-informed, performance-based regulatory system ought not be implemented tfithout the existence of top-level risk-acceptance criteria. The obvious choices for these criteria are the NRC Safety Goal QH0s. As it is the responsibility of the NRC to license individual plants and ensure adequate protection, there seems to be no alternative to plant-specific applications. j RelaWnsbio Between Adecuate Protection and the Safety Goals l

Currently, licensing acceptance criteria are embodied in the concept of " adequate protection." With this concept, a plant that is licensed and complies fully with the applicable rules and regulations, is considered to meet the " adequate protection" standard. " Adequate protection" embodies protection of public health and safety against threats that can be quantified in terms of risk as well as threats, such as sabotage and diversion of special nuclear material, for which the risk cannot now be quantified. In the discussion that follows, the nonquantifiable aspects of adeauate protection are set aside. Since there are many ways in which plants can be designed and operated within the confines of the i regul ations, the natural result is a spectrum of risk levels across the l population of operating plants. This conclusion is consistent with the results of the recent Individual Plant Examination Program. Since each licensed plant 19

3 must, by definition, provide adequate protection, the licensed plant that poses the highest level of risk places a bound on the quantified level of risk to b associated with " adequate protection."

Within the spectrum of risk, it is likely that there are plants with risk levels l above the Safety Goals and other plants with risk levels below. If this is indeed the case, a single risk level that bounds " adequate protection" would be

a risk level greater than the Safety Goal level. For those plants with risk levels 'below the Safety Goals, the difference between the plant risk and the Safety Goals can be viewed as margin. It is from some portion of this margin that plant-specific regulatory relief could be granted. For those plants with

, risk levels greater than the Safety Goals, the challenge will be to eventually

. reduce their risk to below the Safety Goal level within the confines of the l backfit rule.

Reoulatory Transoarency

(

The unquantified " adequate protection" concept is not well understood by the  !

general public becaust the public is unfamiliar with the regulatory process, the )

body of nuclear regulations, and associated underlying technical bases. We  :

believe that a long-term objective of replacing the " adequate protection" concept with a well articulated and quantified " acceptable level of risk" if achievable, i

would enhance the public's understanding and acceptance of the regulatory process

, and would lead to a more uniform level of protection for all individuals living in the vicinity of nuclear plants, i We note that the use of risk-acceptance criteria such as the QH0s will add stability to the regulatory process. This is because the Safety Goals are determined primarily from considerations of societal risk, while the NRC rules and regulations, which are now used to specify adequate protection, change with j time as our understanding of reactor safety issues evolves.

Safety Goals as Risk-AcceotanctCriteria It is our opinion that the QH0s are the appropriate choices for risk-acceptance criteria for plant-specific applications. The Safety Goals are the expression by NRC for "how safe is safe enough." In our opinion, this is what risk-acceptance criteria ought to bE. As we stated in our August 15, 1996 report, the subsidiary CDF goal should be elevated to the status of a fundamental goal.

Elevating the CDF subsidiary goal to the status of a fundamental goal can be considered as a defense-in-depth principle that provides balance between

]

< prevention and mitigation. '

The early fatality QHO generally controls the risks from nuclear plant operations. Our understanding of risk associated with low-power and shutdown operations, or accidents initiated by external events in which emergency response i

i 4

20

c .

f 4 l

l is impeded, is not yet sufficient to draw definitive conclusions concerning the i limiting QHO in these situations.

Additional comments by ACRS Member T. S. Kress are presented below.

l Sincerely,

/ f. f. Ix R. L. Seale Chairman Additional Comments by ACRS Member T. S. Kress While I agree completely with the Committee's report, I think it could be augmented in two respects. First, it could make it clearer that, with respect to plant-specific application of the Safety Goals, we are making two related, somewhat radical proposals - the second more so than the first:

1) That lower tier rick-acceptance criteria (CDF and LERF), now being proposed in Draft Regulatory Guide DG-1061 for use in making decisions regarding requested changes to a licensee's current licensing basis, be derived directly from the prompt fatality QHO and be of such value as to bound all current sites.

2) That, in the long run for enforcement purposes, the prompt fatality QHO be considered as the quantification of a risk level to replace " adequate protection."

Second, guidar.ce on how lower tier criteria are to be derived from the QHO is needed. Consequently, I am including two attachments to these additional comments (one developed by me and a complementary one developed by ACRS Senior Fellow Rick Sherry). These provide examples of how to more rigorously derive the lower-tier criteria. It is suggested that the staff consider these for use if the first proposal above is to be implemented.

Attachments:

1. Kress, T. S., " Risk-Based Regulatory Acceptance Criteria for Plant-Specific Application of Safety Goals," March 1997

2. Sherry, R. R., " Methodology for Estimating Offsite Early Fatality Risk in the Absence of a Level 3 PRA," March 1997

References:

l. Staff Requirements Memorandum dated January 14, 1997, from John C. Hoyle, Secretary, NRC, to John T. Larkins, Executive Director, ACRS,

Subject:

i Meeting with ACRS, 9:30 A.M., Friday, December 6,1996, Commissioners' Conference Room.

21

S

2. Report dated November 18, 1996, from T. S. Kress, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,

Subject:

Plant-Specific Application of Safety Goals.

3. Report dated August 15, 1996, from T. S. Kress, Chairman, ACRS, to Shirley Ann Jackson, Chairman, NRC,

Subject:

Risk-Informed, Performance-Based Regulation and Related Matters.

4. U.S. Nuclear Regulatory Comission, NUREG-1560, Volume 1, Part 1,

" Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance," Sumary Report, Draft Report for Coment, October 1996.

5. U.S. Nuclear Regulatory Comission Draft Regulatory Guide, Draft DG-1061, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," dated February 28, 1997 (Predecisional).

6. U.S. Nuclear Regulatory Commission, Draft Standard Review Plan Chapter 19, Revision L, "Use of Probabilistic Risk Assessment in Plant-Specific, Risk-Informed Decisionmaking: General Guidance," dated March

• 3, 1997 (Predecisional).

I i

l 9

22

ATTACHMENT 1 Risk-Based Regulatory Acceptance Criteria for

! Site Specific Application of Safety Goals T.S. Kress The purpose of this discussion paper is to explore the concept of using the Safety Goal quantitative health effects (OHO) on earfy fatalities to derive lower tier risk acceptance criteria for application on a plant-specific basis. A starting point for expressing the early fatality QHO in t

a form that can be used to derive different tier criteria is the following working definition for the l

risk of early fatalities for any specific plant in terms of the normal determinations of probabilistic j risk assessments (PRAs).

Mean number of early fatalities = { (CDF)a(CCFP)a(C,,)a (1) l where k refers to a specific plant, l l refers to the spectrum of accident aequences, (CDF),.. is the core damage frequency for sequence i of plant k, (CCFP),.. is the conditional containment failure probability for sequence i and plant k,

(C,,),., is the early fatality consequences at site k given the sequence I which has associated with it a source term St,, that may be defined in terms of l

the equivalent release of iodine to the outside environment.

The OHO objective for early fatalities is expressed in terms of individual risk. The Safety Goal Policy Statement specifically states that the early fatality QHO is to be determined by calculating the cumulative individual fatalities within one m!/e of the site boundary, C,,, and dividing that by the population within that same one mile region, P.,. Therefore, Equation 1, for purposes of comparing with the early fatality OHO, should be rewritten as (IR) Individual risk e { (CDF)a(CCFP)a(C,n)p , (2) in order to proceed further, we first note that, in general, the early fatality consequences within one mile of the site boundary can be related to the source term expressed in terms of the equivalent release of iodine by the relationship C,n - (St)"

An appropriate exponent is n = 0.9, which can be obtained from Figure 2 of Attachment 2. l Consequently, if a calculation were available that gave the expected earty fatalities within one mile of the site boundary, C,,,, for any reference source term, St,, then Equation 2 would be rewritten as 4

i 23 l

r

_ _ ._. _ _ _ ~ . _ . . _ _ . _ _ _ . _ _ . _ _ . . _ . . _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _

o ,

l

• 1' lR = { (CDF)a(CCFP),,, (3)

[P,,(St,.'dya)

For our present purposes, Equation 3 can be rewritten as l

(IR) P,, (St,)" l

[(CDF)a(CCFP)a(Stu )" " c (4) i kir The items on the left of Equation 4 are those that are determined by a full-scope Level 2 PRA with source term (to the environment) capability. The items on the right contain the result of a Level 3 consequence analysis for IR and a site characterization parameter, (P ,)(St,)"/C,,,. This parameter can easily be determined using an appropriate ' consequence

• code such as MACCS or CRAC and, in fact, may already exist as part of NUREG/CR-2239, " Technical Guidance for Siting Criteria Development."

NUREG/CR 2239 basically has a characterization of each US site with respect to C,,. The CRAC code was used to determine site-specific early fatality consequences for

- a standard 1120 Mwe plant (for inventory),

- summary evacuation assumptions,

- actual site population and wind rose,

- best-estimate meteorology, and

- a variety of source terms.

Although the values for P., and C.,, out to the one mile boundary were not specifically reported in NUREG/CR 2239, the information may still exist in the archival print-outs of the computed output. In case these data are not currently available, Attachment 2 presents a convenient and robust method for estimating the prompt fatalities at each site for any reference source term and provides an appropriate definition of LERF. '

Direct Comoarison with the OHO Usino Leve12 PRAs with Source Term Caoability

! Equation 4 is an abbreviated form of the early fatality IR that can be used to derive an acceptance criterion for plants with full-scope, Level 2 PRAs with source term capability. For the OHO criterion to be met, the IR must be s 5 x 104/yr. Therefore the acceptance criterion parameter is clearly (5x10 ) P,, (St,)"

[(CDF)a(CCFP)a(Sta)" s (5) )

kir l

The criterion, Equation 5, can also be cast in terms of a conditional early release probability criterion, CERP, that can be defined as l

l

'(((CDF)al x CERP a [(CDF)a(CCFP)a(S'u)" l 24

. _ _ _ _ . . ..-_-- -. ._. .. . _- _ . . . ~~ - . . - . - . . - - _ _ . - - -

Or  !

l

[ (CDF),,(CCFP),,(St,,)"

(ggp# ,

[(CDF),j in terms of acceptance criteria similar to what was recently proposed by the staff with coordinates of CDF and LERF with a basic tenet that CDF should be s 1 x 10d/yr, the j appropriate plant specFic CERP acceptance criterion would be l

(5x10-7)(St,)"(P,,)

i

' CERPu s {

(10")C,,,

l if a single CERP, is desired that would bound all current sites, the site with minimum P.,/C.,,

would have to be chosen.

Level 2 PRA Without Source Term Caoability l

For the case of a plant that has Level 2 PRA capability to determine CCFP but not to establish the various associated sequence source terms, it is desired to develop a lower tier criterion in terms of a defined CCFP that would be bounding with respect to ensuring that the early fatality OHO is met on a plant specific basis.

To do this from Equation 5, it will be necessary to use a single source term that is sure to bound each sequence source term. A possible value that would be bounding (for PWRs only) is 0.5 (release fraction of iodine equivalent). With this value, Equation 5 as a PWR plant specific bounding criterion becomes

[(CDF)a(CCFP),, s (5x10-7)(P,,)(St')"

(C,,,) (.5)"

The parameter on the left is what I believe the statt used in hs decision chart and called a 'large early release frequency" (LERF). This would more aptly be called a containment failure frequency.

l To express this in terms of a conditional containment failure probability, we could use the usual definition of CCFP.

CCFP# n " "

[(CDF)g f

25 i

i

l . .

Th2n, d

cn acceptince value (for PWRs only) for use in a ' decision" chIrt along with CDF =  ;

10 /yr becomes CCFP s (5x10~7) (P,,) (St,)*

u (10") (Caf,) (.5)"

A separate acceptance criterion could be developed for BWRs using a bounding St that would have to recognize the effects of the suppression pool and its bypass. An optional way to account for the differences between BWRs and PWRs could be to make different choices for the sequences to go into calculating the CCFP. This is what the staff chose to do - thereby preserving the same acceptance criterion for both reactor types.

For site bounding values, the minimum P,3/C,,, for each reactor type would be chosen. It would l .

be this bounding value that should, in the case of PWRs, be compared with the CCFP = 0.1 l that is often mentioned as an appropriate criterion to use along with CDF = 1 x 10d /yr. I I

l l

I l l l

! 26 l

l l

1 ATTAOMENT 2 l

l Methodology for Estimating Offsite Early Fatality Risk in the Absence of a Level 3 PRA Rick Sherry- Senior Fellow Advisory Committee oc Reactor Safegua-ds Introduction This paper defines a simple relationship between the Safety Goal Quantitative Health Objective (QHO) for Individual Early Fatality Frequency (IEFF) and a Large, Early Release Frequency (LERF) that can be used to estimate the Safety Goal QHO for a specific plant at a specific site.

This paper also provides a quantitative definition of the LERF. The relationship utilizes simple site-specific characteristics (wind direction frequencies and population demographics) and results from a Level 2 plant-specific probabilistic risk assessment (PRA) (release category frequencies l

and source term characteristics). 1 l Estimates using this methodology have been compared with results from detailed calculations performed using the MELCOR Accident Consequence Code System (MACCS)(Ref.1) for the NUREG-1150 study (Ref. 2), the Grand Gulf Shutdown study (Ref. 3) and the Maine Yankee PRA/IPE. This comparison includes internally initiated sequences, seismic sequences, and shutdown sequences.

l Summary of Methodology l

The relationship between the IEFF Safety Goal QHO and the LERF is defined as:

IEFF = LERF x El Equation I l

l 16 1 [ P, x F, where: El = Exposure index = Equation 2

{ P, F, = the relative frequency wind blows toward sector i P, = population in sector I within one mile of the plant i

27 l

l N,,

and: LERF = [RC,(: early I > 10% and Evacuation Delayed) x (1 - F,,,,,,,)

a N,,

+ { RC,(; *otal I > 10%) x F,,.,,,, Equation 3 i & *)

RC, = rhe frequency of release category k N,, = number of release categories l

l F,,.,,,, = rhe population faction not evacuating

( .

l I Discussion l The Exposure Index (EI) provides a measure of the average probability that a specific individual within one mile of the plant would be exposed to a lethal radiation dose (given that a release occurs from the plant of sufficient magnitude to produce lethal doses) and assuming that the individual does not evacuate. The exposure index defined above is a slight variation from that initially developed for the Generic Environmental Impact Statement for License Renewal of Nuclear Plants (Ref. 4). Table I shows the calculated values for the exposure index for a number of nuclear plant sites using site specific wind direction frequency and population data. Note that all calculated values of the El fall well within a factor of two of the nominal value of 1/16 which would result if the population within one mile of the plant and the wind direction frequency were uniformly distributed in the sixteen w:ind direction sectors.

In the above definitions, the release categories which are included in the summations are either 1) those with an early release fraction ofiodine greater than 10% of the core inventory and for which evacuation is delayed or 2) release categories where the total iodine release fraction (sum l l of early and later releases) exceeds 10% and evacuation does not occur. The second summation accounts for the fraction ofindividuals that do not evacuate and are assumed to be exposed to all time phases of the radionuclide release. This second summation also governs severe seismic sequences where evacuation is assumed to be ineffective.

A ten percent iodine release fraction has been selected as the threshold for source term magnitudes which can result in lethal doses. The selection of 10% was based principally on the work of Kaiser (Ref. 5) which indicated that a threshold for early fatalities occurs at a release fraction of approximately 10% of volatile fission product (I, Cs, Te) release fraction. Figure 1 from Kaiser illustrates this behavior.

An early release is defined as the release which occurs at the time of contamment failure (assuming core damage has occurred prior to containment failure). For sequences where l containment integrity has been lost prior to core damage, early release beghts when core damage i

i 28 i

l

l . .

i commences. This definition for early release is identical to that provided in NUREG-1150. l Typical periods of release duration for early release in the NUREG-1150 study are from several I minutes to several hours.

l Regardless of the magnitude of the source term, if evacuation commences sufficiently prior to the time when the release ofradionuclides begins then the probability of early fatalities is l

dramatically reduced. In the NUREG-1150 study, three release category timing subgroups were defined for each release category. For sut> group 1,it was assumed that evacuation commenced at i least 30 minutes prior to the start of radionuclide release. For subgroup 3, the .: tart of evacuation was delayed until one hour or more after radionuclide release had begun. For subgroup 2, the '

start of evacuation was assumed to occur within a time window from 30 minutes before, to one ,

bour after, the start ofrelease.

l l Figure 2 illustrates the impact of these various evacuation assumptions on the Conditional Individual Early Fatality Probability' (CIEFP). This figure plots the CIEFP against the iodine [

i release fraction. Individual data points for the three release category subgroups are shown with  !

different symbols. This figure illustrates the effectiveness of early evacuation in reducing the Individual Early Fatality Risk. The diamond shaped symbols represent sequences for which evacuation was delayed until one hour or later after the start of radionuclide release (subgroup 3).

For these sequences, the peak CIEFP is on the order of 3 x (10)4. The triangle shaped symbols i represent sequences for which evacuation was initiated at least 30 minutes prior to the start of the release of radionuclides (subgroup 1). These sequences have a CIEFP of order 2 x (10)" or less. l l These results are dominated by the fraction of the affected population who are assumed not to l

l evacuate. For sequences characterized by evacuation commencing at about the same time as the l

l start of radionuclide release (subgroup 2), the results (shown as circles) generally fall between the results for subgroups 1 and 3. l Equation I can be rearranged and used to estimate a plant specific LERF subsidiary objective:

LERF = Equation 4 l EI This equation along with the plant specific Exposure Index values shown in Table I have been used to calculate a plant specific LERF objective. These results are also shown in Table 1. Note t

' The CIEFP is conditional on the occurrence of sequences (release categories) with

! the indicated volatile fission product release magnitude.

f l l t .

[ 29

that for these calculations, the assumed value for the Safety Goal QHO IEFF is taken as 3 x (10)'

per year.2

! Eaample Application Results

'ne simple methodology has been applied to six cases. These cases are:  ;

Surry Internal Event Initiated Sequences (NUR3G-1150) l Surry Seismic Sequences (NUREG-1150)

Peach Bottom Intemal Event Initiated Sequences (NUREG-1150)

Peach Bottom Seismic Sequences (NUREG 1150) l' Grand Gulf Shutdown Sequences (NUREG/CR-6143)(Ref. 3)

Maine Yankee Intemally Initiated Sequences (PRA/IPE subndttal) l' The results of the example application are summarized in Figure 3. In this figure, the Individual Early Fatality Risk Frequency calculated using the simple methodology is compared with the mean and 95th percentile values from the PRA offsite consequence (MACCS) code calculations. l l

In all cases, the simple methodology produces estimates for the IEFF that are above the PRA mean values. For the two seismic cases, the results fall between the PRA mean and 95th  :

l percentile values, and in the remaining cases the results lie above the 95th percentile (but within a factor of three of the 95th percentile value).

The limited comparison summarized in Figure 3 indicates that this methodology may provide a simple and easy to use approach for providing reasonably robust estimates for the IEFF for PRA analyses lacking a detailed Level 3 offsite consequence. This methodology has been applied to a

! . broad spectrum of PRA sequence classes and in all cases the comparison with the PRA results

[ ' have been favorable.

Discussion of Results Intemallv Initiated Seauences (NUREG-1150)

For Peach Bottom intemally initiated sequences, all release categories that met the criteria for early iodine release greater than 10% also had evacuation beginning prior to the start of release.

This is the principal reason that the MACCS calculated IEFF is low for Peach Bottom. It is only the small fraction of the population that does not evacuate which contributes to the calculated 2

Based on data from the National Center for Health Statistics the total U.S. societal l .

accident fatality rate was 34.4 and 34.6 deaths per 100,000 individuals in 1993 and 1994, respectively. ' Ibis translates into an individual early fatality Safety Goal QHO of 3 x (10)'per reactor year.

1 4

I 30 i

i IEFF. Consequently, in this study the estimated IEFF for Peach Bottom is controlled by the second summation term in Equation 3 (as for the NUREG 1150 MACCS analyses it was assumed that 0.5% of the population did not evacuate and Fm was set to this value).

For the Surry plant, the dominant contributors to a large early fission product release and early fatalities for intemally initiated sequences were containment bypass events (12% of CDF). These l sequences contribute about 80% of the early fatality risk. For the fast developing interfacing .

l system LOCA bypass sequences, warning times were typically withm one hour of the start of radionuclide release. Consequently, evacuation would generally begin after the start of release.

Seiemic Secuences (NUREG-1150)

The NUREG-1150 results for seismic sequen:es for the Peach Bottom and Surry plants using the 1

LLNL seismic hazard curves were compared for this study. The NUREG 1150 seismic sequence consequence analyses were separated into two categories - low Peak Ground Acceleration (PGA)(< 0.6 g) and high PGA (> 0.6 g). For seismic sequences, the offsite emergency protective action assumptions input into the MACCS analyses differed from that for internally initiated events. Particularly important was the assumption that for the high PGA category sequences evacuation was assumed to be ineffective (i.e., does not occur). Consequently, for this study the total iodine release fraction for high PGA seismic sequence classes was used (i.e. the second summation in Equatica 3 with F_ set to one).

Grand Gulf Shutdown Seouences (NUREG/CR-6143) -

i For shutdown sequences, the activity level ofimportant volatile 5ssion product species will be reduced by decay from their full power levels and this reduction should be accounted for in analyses. However, for this application this was not done. Even with this limitation the estimated IEFF using the simple methodology was in good agreement with the MACCS results. As for the Peach Bottom intemally initiated sequences, warning times for Grand Gulf shutdown sequences were sufficiently early that evacuation commenced prior to the beginning of fission product release, and it is only that small fraction of the population that does not evacuate which contributes to the calculated IEFF. Consequently, the estimated IEFF is controlled by the second summation term in Equation 3.

Maine Yankee Internally Initiated Seanences (PRA/IPE)

The IEFF for Maine Yankee is dominated by sequences where containment failure occurs as a result of a hydrogen combustion event near the time of reactor vessel meltthrough. This type of failure was predicted to occur for about 8% of the core damage frequency. For those sequences contributing to the IEFF, the radionuclide release is predicted to begin about one hour after a general emergency would be expected to be declared. In the Maine Yankee PRA, delay times 31 I

ranging from 1 to 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> were assumed (dependent on the time of year). Hence, for these sequences, radionuclide release generally begins prior to, or coincident with, the start of evacuation.

l Enhancements - Use ofIodine Equivalent 1 Although this paper has shown that reasonably good correlation exists between the IEFF and the LERF calculated using Equation 3 (which uses an iodine release fraction of 10% u threshold for eariv fatalities), it was also observed that under certain conditions other fission product species grey, em make significant contributions to the early fatality risk. It was observed that if the Ru or Lai i;roup release fractions exceeded about 1 to 5% of core inventory, they began to make a significant contribution to early fatality risk. Furthermore, as discussed above, for sequences ,

initiated during shutdown which may have significant decay periods, the contribution ofiodine 1 isotopes to the early fatality risk will be significantly reduced and other radionuclide groups will begin to dominate. For these reasons, a parameter that can be evaluated from the release fraction for all radionuclide groups (and as a function of shutdown time)is desirable. An example of the ,

type ofiodine equivalent parameter that is required is presented in NUREG/CR-5164 (Ref. 6)  !

wl ich is based on work reported in Reference 7. In these studies, weighting factors for the I contribution to early fatality risk fer each radionuclide species group were developed. These  ;

weighting factors (which are multiplied by the radionuclide group release fraction) can be used to develop a total iodine equivalent release. ,

References  :

1. Chanin, D. l., et al.,"MELCOR Accident Consequence Code System (MACCS): User's i Guide," NUREG/CR-4691,Vol .1, Sandia National Laboratories, February 1990.

2. " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"NUREG-1150, December 1990. [ Note, references to NUREG-1150 also imply reference to the NUREG-1150 supponing documents including applicable volumes of NUREG/CR-4551]

3. Brown T. D. et al., " Evaluation of Potential Severe Accidents Dunng Low Power and i Shutdown Operations at Grand Gulf, Unit 1," NUREG/CR-6143, Vol. 6, Part 1, March 1995. '

' 4. " Generic Environmental Impact Statement for License Renewal of Nuclear Plants",

NUREG-1437, Vol.1, May 1996.

5. G.D. Kaiser, "The implications of Reduced Source Terms for Ex-Plant Consequence -

Modeling," ANS Executive Conference on the Rami 6 cations of the Source Term, Charleston, SC, March 12,1985. j

6. Madni 1. K., et al., "A Simplified Model for Calculeing Early Offsite Consequences from ,

t

?

32

Nuclear Reactor Accidents'" NUREG/CR-5164, Brookhavcn National Laboratory, July I 1988.

7. Alpen J., et al," Relative Importance ofIndividual Elements to Reactor Accident Consequences Assuming Equal Release Fractions," NUREG/CR-4467, Sandia National Laboratories, March 1986.

l l

5. Brown T. D. et al., " Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Grand Gulf, Unit 1," NUREG/CR-6143, Vol. 6, Part 1, March 1995.

l l

l l

l l

i l

l 2

33

Table 1

, Example of Plant Specific Exposure Index and

! Derived Subsidiary LERF Objective l

I Plant EI Derived Subsidiary I

LERF Objective Grand Gulf .065 SF4 Surry .077 4E4 Sequoyah .045 7E4 Maine Yankee .040 SE-6 l Zion .081 4E-6 Peach Bottom .075 4E-6 Lasalle .083 4E4 Nominal .063 (1/16) SE-6 I

l 34

j 1

l

)

i i

I 1

1 I

w', .

I 4saweim., w .

so 8 - .

. i s I* *.

! s*

1 ,e2 -

.I 1 '.

so e' . .

8

- t 3 1, e s'.

e .

l l

t i l i

u t

l ,

8 8

8 4 44 _ , 3_ , 4 44 ee . 6 4 4. .  !'

,,a ,,-s ,o-2 ,- i ,

a.=ress w i. c. .w ve =====

{

Figure 1. Early Fatalities i

i l

l l

l l

l

{

(

l 35

Figure 2 Surry Individual Early Fatatality Risk InternalInitiators 1.00E+00 _

p--;jQ  :. .

- -- .- 7 , :- ,

--.- 4- - .

_ _ '[_

1.00E-01 - - : m- .=-_ .. .

- ::: .- = -

-= -

- .. _ . - . -- ~

, .~ '.

'.~

.n e . - .

m e.

- L.

$.00E=0E - .-.-.--~~--' -~

, o 1 a -.:.- : . ._:-~~--.:.~-. - - q* -

: = :: =

g . . - -

e w .:

-e , , .  :. .e g . - . . _ . . .- .._. - -

.e.-

w . .- . -.

= _ - . _

W.00E43- =_r g :=  : _. --

• e Evac Acer

~ ~ - c. - --

eEvoc Dwing 5.i

'~

i.'

. . __ . . _T -

a Evac Price

%..- -.-m.s. ..e . .. . _ . --

.6 a

. .00E44 - - : :- ^

x "~. .....- .. . . . ^ - -

E

. :. '?? .~ i- _. .

~, ~

o - - - --  !

u -. . ._

9 , . . . .. - '

1.00E 05 - . - $ -- , . * -

. .e

~

e

' 1.00E 06 1.00E-83 1.00E-02 1.00E-01 1.00E+00

. T lodne Release Fraction

..__.___________.-__._-.._____.___m__ - _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _

Figure 3 IndMdualEstfy Fatality Risk PRA (NUREG-1150) v Simple Est'anetor 100E44 i

I

, nog.oS . . _ _ . _ _ _ _ . . _ . _ _ _ _ . _ __

JL 4>

II 100E 06 --

---

at-- -- - --

o ll 100E-07 -- --- -- -- --- - - - - - - - - -- - - - - - -- --

N .

ePRA m APRA-95th w II N OLERF-Est '

1.00E-ce - - - - - - - - - - - - - - - - - - - -- - - -- - - - - - ~~ ~~~

11 1.00E-09 - -- - - - - - - - - - - - - - - --- --- - - - -- -

o 4>

JL 100E-10 -- -- - - - - - - - - - - - -

ll 100E ti Suny Suny peach Peach Grand Maine Intemal Seismic notsom Bottom Gulf Yankee (I150) (I150) Intemal Seismic Shutdown Internal  ;

(l150) (I150) '

(IPF1PRA)

--, em .- s a -

L-a,-p1-.n m A , _ _ > _ - n -.- --~.. -u== ,- - A. , - ,. ..J.. man,__s-- n- % -,eA4 4 4 l

I ITEM B.2:

PROPOSED REGULATORY APPROACH ASSOCIATED WITH STEAM GENERATOR INTEGRITY (DR. SEALE) 1 l

l l

l ITEM B.2: PROPOSED REGULATORY APPROACH ASSOCIATED WITH STEAM l GENERATOR INTEGRITY l

i On September 12,1994, based on its review of the draft generic letter concerning the proposed i voltage-based repair criteria for Westinghouse steam generator tubes, the Committee provided )

a report to the Commission. In this report, the Committee agreed with the staff position that I rulemaking was the preferred regulatory approach to the problem of steam generator tube I degradation, but was skeptical that a new rule could be developed as expediously as specified by the proposed staff schedule.

The Committee reviewed the proposed steam generator rule and an associated regulatory guide .

I and provided a letter to the Executive Director for Operations (EDO) on November 20,1996.

Some of the comments and recommendations in this letter included the following:

• In its present form, the rule is a performance-based regulation almost completely ~

divorced from any direct relation to risk objectives. Such a performance-based mle proliferates the incoherence problems of the present deterministic approach.

• To be risk informed and performance based, the regulatory guide should begin with i a clear statement of its objectives, followed by a statement of the performance criteria and the guidelines for meeting the criteria.

• The Committee agrees that the staff should approve the performance criteria that are proposed by licensees to implement the steam generator rule. Industry, however, should be provided more flexibility to propose alternative performance criteria supported by an appropriate risk analysis.

• The staff position is that the regulatory guide provides sufficient guidance for developing an acceptable methodology and that formal review of industry-developed repair criteria and procedures will not be required.

• The staff should prepare a point-by-point response to the outstanding issues in the differing professional opinion, and resolve generic safety issue GSI-163, " Multiple Steam Generator Tube Leakage," before implementing the steam generator integrity rule.

The EDO responded to the above ACRS letter on January 2,1997. In this letter, the EDO suggested holding additional meetings to discuss the Committee comments and recommendations. Following a meeting between the Committee and the staff on January 9, 1997, the Executive Director for the ACRS issued a memorandum to the staff on January 31, 1997, as suggested by the Committee, stating that the ACRS request that the staff, at the next ACRS Subcommittee meeting, respond to the comments and recommendations included in

, the November 20,1996 ACRS letter and to several concerns raised by individual Committee members. Some of the individual member's concerns are as follows:

i

! 38 1

j l

The performance criteria, which are intended to ensure that the risk due to thermally induced tube failures during severe accidents is acceptable, appear to introduce new ,

requirements. The basis for the introduction of these requirements is unclear. Very l little explanation has been provided that clearly identifies how the performance criteria and program requirements are related to risk analyses.

The staff should clarify whether performance criteria are derived from top level risk requirements or from defense-in-depth considerations. The staff should identify the number of and bases for the defense-in-depth criteria.

The regulatory guide does not describe, in all cases, a standard for complying with the performance criteria or program requirements.

The regulatory guide does not, in all cases, identify how the licensees can demonstrate that reasonable assurance has been achieved.

l l

The ACRS Subcommittees on Materials & Metallurgy and on Severe Accidents held a jomt '

meeting on March 4-5,1997, with the NRC staff to discuss the draft risk assessment of severe accident induced steam generator tube ruptures, and the associated draft regulatory analysis.

The staff stated that based on the results of these two efforts, it was reassessing its regulatory l approach for addressing steam generator tube integrity issues. Specifically, the staff is reexamining alternative regulatory approaches for addressing steam generator tube integrity. l These include utilizing the current regulations.and the PRA Implementation plan as a framework for regulating steam generator tube integrity issues rather than a new rule.

The staff is drafting a memorandum to the Commission that will discuss the general conclusions from the draft risk assessment, the major conclusions from the draft regulatory analysis, the implications for steam generator rulemaking, and an alternative regulatory approach to accomplish the original objectives of the steam generator mle. The Committee plans to review and comment on the staff approach after receiving the staff memorandum.

Attachments:

• Letter dated November 20,1996, from T. S. Kress, Chairman, ACRS, to James M.

Taylor, Executive Director for Operations, NRC,

Subject:

Proposed Rule on Steam Generator Integrity (pp. 41-45)

• Letter dated January 2,1997, from James M. Taylor, Executive Director for Operations, NRC, to Thomas S. Kress, Chairman, ACRS,

Subject:

Staff Response to ACRS Comments on Proposed Steam Generator Rule (pp. 46-47) l 39

• Memorandum dated January 31, 1997, from John T. Larkins, Executive Director, ACRS, to Ashok C. Thadani, Office of Nuclear Reactor Regulation,

Subject:

Steam  !

Generator Integrity Rulemaking (pp. 48-50) i t

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UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555

\ ces,*

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November 20, 1996 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ,

SUBJECT:

PROPOSED RULE ON STEAM GENERATOR INTEGRITY During the 436th meeting of the Advisory Committee on Reactor Safeguards, November 7-9, 1996, we reviewed the technical bases for the proposed steam generator integrity rule and an associated regulatory guide. During the 432nd meeting of the ACRS, June 12- )

14, 1996, and meetings of the Joint Subcommittees on Materials &

Metallurgy and on Severe Accidents, June 3-4 and November 5-6,  ;

1996, we heard presentations on subjects related to this matter.

During these reviews, we had the benefit of discussions with representatives of the staf f, the Nuclear Energy Institute, and the Electric Power Research Institute, as well as the author of a differing professional opinion. We also had the benefit of the documents referenced.

The proposed steam generator integrity rule is intended to provide a risk-informed and performance-based regulation to replace an axisting prescriptive regulation. In its present form, the rule is o performance-based regulation almost completely divorced from any direct relation to risk objectives. Such a performance-based rule  ;

proliferates the incoherence problems of the pusent deterministic cpproach. The proposed rule preserves a tenuous connection between l

" design-basis space" and " risk space" without clearly articulating the risk objectives.

Some of the characteristics exhibited in the development process of the rule and regulatory guide include the following:

o dif ficulty in reaching agreement on the performance criteria, o incomplete and sometimes perfunctory analyses required to l

provide an assessment of rulative risk, l 0 reliance on core-damage frequency alone as an indicator of risk, and 41

l i l

Jcmno M. Tcylor -2 -

e recourse to defense-in-depth without specific criteria for its use.

We believe that more direct consideration of risk could have '

avoided some of these difficulties.

A controversial element of the proposed rule and regulatory guide is the introduction of severe accident issues into an area that has been exclusively resolved by using a design-basis analysis. This l cxtension of the scope of accident analysis is necessary to make l risk-informed regulatory decisions and is part of the cost of moving toward risk-informed regulation. Since licensees have done risk-informed analyses for the Individual Plant Examination (IPE) process, we believe that the analysis for addressing severe occident events should not be overly burdensome to them.

! Steam generator tube ruptures are small contributors to the total l core-damage frequency, but may be risk significant due to l containment bypass effects. In previous analyses, the staff performed limited assessments of primary side fission product attenuation and neglected secondary side attenuation. The regulatory guide now proposes that the licensees deal with the risk

, of a thermally induced tube f ailure either by demonstrating that I the frequency of the initiating events is sufficiently low (10-'/ reactor year) or by demonstrating that the conditional probability of tube failure, given that an initiating event has occurred, is low (on the order of 0.1). We believe that licensees l ohould also be given the option to demonstrate that, even if i thermally induced tube ruptures occur, the associated risk is low when a more realistic treatment of fission product attenuation is made.

i We are concerned that the proposed regulatory guide, as presented, i could send the wrong message to licensees that risk-informed and performance-based requirements are add-ons to the traditional design-basis accident approach and can only result in an additional burden. We believe that to be risk informed and performance based, the regulatory guide should begin with a clear statement of its l objectives, followed by a statement of the performance criteria and l the guidelines for meeting the criteria. We note that the staff has stated that the proposed performance criteria have been derived from risk analyses, but we have not seen these analyses. Rewriting the regulatory guide is not a trivial task, but could result in a l regulatory framework that could be used as a model for future risk-l informed and performance-based rulemaking efforts.

In other applications of performance-based regulation such as the Maintenance Rule, the licensees have been permitted to determine cppropriate performance criteria and have been given more flexibility in developing the methodology used to determine whether the criteria have been met. For the steam generator rule, the 42 l

James M. Taylor I staff has concluded that it should approve the performance criteria  !

that are proposed by licensees to implement the steam generator rule. We agree with the decision of the staff that it should cpprove the criteria. Industry, however, should be provided more i flexibility to propose alternative performance criteria supported l by an appropriate risk analysis. We would like to review all of l

l the supporting documentation before commenting on the specific criteria that have been proposed in the regulatory guide.

The demonstration that the criteria have actually been satisfied l requires a complex process of nondestructive examination and i evalui. tion of structural integrity and leakage during operation and l

design-basis accidents. The methodology required for these evaluations is not well established. Thus, the staff has felt constrained to provide a great deal of detail in the proposed regulatory guide to describe the characteristics of an acceptable methodology. Although we are not yet prepared to endorse the regulatory guide, we believe that the present immaturity of the methodology and the importance of the results justify such an i approach. '

l ,

The staf f posit. ion is that the regulatory guide provides sufficient l guidance for developing an acceptable methodology and that formal l review of industry-developed repair criteria and procedures will  ;

not be required. We would like to review the results of a " trade '

study" of the preapproval approach vs. the post-implementation inspection approach to methodology acceptance.

Industry has questioned whether safety factors proposed in the steam generator rule are more conservative than those required by  !

the ASME code. We encourage the staff to consider the industry's '

l nrguments.

l Industry accepts the performance criterion proposed by the staff for primary-to-secondary leakage. Industry stated that this

leakage criterion ought not be ipso facto a trigger for inspection l or enforcement of regulations concerning the steam generator rule.

This is a valid concern. Excessive leakage does not necessarily indicate a failure of the steam generator program. Adequate opportunities for staff action are available if failures of the program are discovered following a plant shutdown due to excessive primary-to-secondary leakage.

We are looking forward to reviewing the staff NUREG report concerning the staf f

• s treatment of thermally induced tube f ailure.

We are especially interested in the treatment of elevated temperatures resulting from flow through leaking tubes, and i coupling between aerosol deposition and thermal hydraulics. I A dif fering professional opinion (DPO) was filed on July 11, 1994.

We have reviewed the contentions in that DPO and summarized them in l I 43 l _ __ _ _ _ _ _ ._

Jcm3D M. Toylor -4 -

the attachment. We also note that Generic Safety Issue (GSI)-163,

" Multiple Steam Generator Tube Leakage," identified in 1992 has yet to be prioritized and resolved. Both the DPO and the GSI are directly related to the proposed rulemaking. We urge the staff to prepare a point-by-point response to the issues in the DPO and to prioritize and resolve GSI-163 before implementing the steam generator integrity rule.

Dr. William J. Shack did not participate in the Committee's deliberations regarding this matter.

I Sincerely,

, f.

T. S. Kress Chairman

Attachment:

Summary of Differing Professional Opinion Issues - Presented to the ACRS on November 7, 1996

References:

1. Memorandum dated October 25, 1996, from Brian Sheron, Office of Nuclear Reactor Regulation, to John Larkins, Executive Director, ACRS,

Subject:

ACRS Review of the Proposed Steam Generator Rule (forwarding the proposed steam generator rule and draft steam generator regulatory guide)

2. Memorandum dated May 1, 1996, from James M. Taylor, Executive Director for Operations, NRC, to Joram Hopenfeld, Office of Nuclear Regulatory Research, NRC,

Subject:

Resolution of Dif fering Professional Opinion Regarding Voltage-Based Repair Criteria for Steam Generator Tubes, dated July 13, 1994

3. Memorandum dated July 15, 1994, from James M. Taylor, Executive Director for Operations, NRC, to John T. Larkins, Executive Director, ACRS,

Subject:

ACRS Review Of Proposed Generic Letter 94-XX, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes [ forwarding Differing Professional Opinion]

4. Report dated September 12, 1994, from T. S. Kress, Chairman, ACRS, to Ivan Selin, Chairman, NRC,

Subject:

Proposed Generic Letter 94-XX, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes"

5. Memorandum dated September 30, 1994, from Joram Hopenfeld, Office of Nuclear Regulatory Research, NRC, to John T.

Larkins, Executive Director, ACRS,

Subject:

Comments On ACRS l Review Of Generic Letter " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes" l

44

-. . . .. . . ~ - - - .- - -..- . - . . . - .. - _ - - -

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SUMMARY

OF DIFFERING PROFESSIONAL OPINION ISSUES PRESENTED TO THE ACRS ON NOVEMBER 7, 1996 l

The DPO author estimates core-damage frequency with containment bypass to se 10 -' -

3.4 x 10-* events / year. He stated that the uncertainties associated with characterizing steam generator tube defects and severe accident phenomena are not sufficiently understood to properly model tube rupture events. Tubes may fail before the surge line due to:

e crack networking and characterization of flaws not being adequately determined by nondestructive examinations, e increased heat transfer caused by flow through tube cracks, e- cracks in tubes opening due to increased pressure, e cracks in tubes unplugging at elevated pressure, and l

e jets from tube cracks eroding adjacent tubes.

The DPO author stated that the staff should document the i assumptions and models used to study hidden uncertainties.

l 1

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, . ... . - - . _ . . . ~ _ . . . . . - - . . . - - - . . . . . . -

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\ UNITED STATES g NUCLEAR REGULATORY COMMISSION

• CASHINSTCN, D.C. SAAH001

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January 2,1997 Dr. Thomas S. Kress, Chairman -

Advisory Comittee on Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

STAFF RESPONSE TO ACRS CDPMENTS ON PROPOSED STEAM GENERATOR RULE l

f

Dear Dr. Kress:

I am responding to your letter of November 20, 1996, in which you forwarded

the comments of the Advisory Comittee on Reactor Safeguards (ACRS) on the

i. technical bases for the proposed steam generator rule and an associated draft regulatory guide. The staff appreciated the opportunity to brief the ACRS on this matter in November 1996, and the timely response from the committee.

l The staff has concluded that additional meetings on the proposed rulemaking are appropriate. In this regard, NRR and ACRS staff are working to schedule several additional meetings. It is our intent that the first of these meetings, to be scheduled in the near future, would provide an opportunity to l discuss some broad issues related to imp 1.ementation of risk-informed

! regul ation. These issues are related not only to the proposed steam generator rule, but also to other risk-informed initiatives that the staff has recently presented to the ACRS. Following the next meeting, we will schedule additional subcomittee and full comittee meetings to further discuss the specifics of the proposed steam generator rulemaking.

I would also like to note that it is the staff's intent that the differing professional opinion'(DPO) concerns that were presented to the ACRS and sumarized in your letter, will be addressed in the rulemaking package.

l l

l CONTACT: Timothy A. Reed, NRR

415-1462 1

f 46

l Dr. Thomas S Kress ' I appreciate your cooperation with the staff in resolving these issues so that the agency can proceed with the proposed steam generator rulemaking.

! Sincerely, l l

.'iyor '

E ecutive irector for Operations cc: Chairman Jackson Commissioner Rogers Commissioner Dicus 1 Commissioner Diaz l Commissioner McGaffigan SECY 1

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1. 'g UNITED STATES 8 n NUCLEAR REGULATORY COMMISSION

$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O A WASHINGTON D. C. 20555

• e..* f January 31, 1997 MEMORANDUM To: Ashok C. Thadani, Acting Deputy Director

, Office of Nucl R e lation l

l FROM:

John T. Larkins, txec ive.f'an L Director Advisory Committee on Reactor Safeguards

SUBJECT:

STEAM GENERATOR INTEGRITY RULEMAKING

During the joint aseting of the ACRS Subcommittees on Materials &

, Metallurgy and Severe Accidents on January 9, 1997, the staff discussed the broad-issues related to risk-informed, performance-based regulation, as well as the status of developing the steam generator integrity rule, the associated regulatory guide, and the l technical analyses. This meeting was also intended to discuss the l staff response to ACRS comments and recommendations contained in l its November 20, 1996 letter to the Executive Director for i operations concerning the proposed steam generator integrity rule l and the associated regulatory guide.

In fact, the January 9 meeting was primarily devoted to a discus-cion of the staff's approach to completing the Standard Review Plan cactions and supporting documentation associated with risk-informed, performance-based regulation. In addition, a discussion was held on the approach for developing the steam generator integrity rule, the related regulatory guide, and other documents.

Although these discussions were somewhat helpful to the members, l this meeting did not completely resolve the comments and recommen-dations included in the November 20, 1996 ACRS. letter. During the Subcommittee meeting, the staff expressed an intent to incorporate come of the comments and recommendations into the proposed

! documents; however, the extent to which the ACRS concerns would be reflected in these documents was not clear.

The meeting did allow the members to sharpen the focus of some of their concerns. Many of the concerns of the members are about how the specific performance criteria and guidance described in the regulatory guide relate to risk. Much of the difficulty that the IIembers have had in reviewing the steam generator integrity rule cnd regulatory guide is due to the unavailability of written documents (e.g., NUREG and regulatory analyses) that describe the bases for the performance criteria. The information provided in

)

general terms by the staff during subcommittee meetings was I -

48

l . .

2 insufficient for the members to clearly understand the bases for cuch criteria. The Committee requests that the staff, at the next Subcommittee meeting, respond to the ACRS comments and recommenda-tions included in the November 20, 1996 letter and to the following concerns raised by individual Subcommittee members:

• The performance criteria, which are intended to ensure that the risk due to thermally induced tube failures during severe accidents is acceptable, appear to introduce new requirements.

The basis for the introduction of these requirements is unclear. Very little explanation has been provided that clearly identifies how the performance criteria and program requirements are related to risk analyses.

• The staff should explain how the proposed. rule will transition from the existing deterministic regulatory process, which is l

i based on design-basis accident information, to a risk-in-i formed, performance-based process, which is based on risk I information.

• The regulatory guide does not clearly state the objectives, l

i functional requirements, performance requirements, verifica-tions, acceptable solutions, and alternative solutions associated with the performance criteria.

l l

• The staff should clarify whether performance criteria are l derived from top level risk requirements or from defense-in-depth considerations. The staff should identify the number of and bases for the defense-in-depth criteria.

• The regulatory guide does not describe, in all cases, a standard for complying with the performance criteria or program requirements.

• The regulatory guide does not, in all cases, identify how the licensees can demonstrate that reasonable assurance has been achieved.

• The staff should explain the basis for the 0.05 tube failure per year criteria -- in particular, how the value for steam  ;

generator tube ruptures is an appropriate allocation of the total conditional containment failure probability; how the criteria for tube plugging are derived from inspection findings, leak rates, or voltage criteria; or how plugging criteria are derived from a probability of tube failure.

49

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• The regulatory guide does not explain how licensees should )

demonstrate that the spontaneous probability of steam genera- '

, tor tube failures will be below the assumed failure criteria.

1

• The staff should explain the basis for the allocation of 20 percent of the tube failure probability criteria to each l degradation mechanism. 1 l

i The Committee looks forward to meeting with you and your staff in the near future. If you have any questions or need additional information, please contact Noel Dudley of my staff at 415-6888.

\

cc: ACRS Members  !

J. Hoyle, SECY l D. Morrison, RES l

J. Mitchell, OEDO F. Miraglia, NRR

.l B. Sheron, NRP J. Cortez, RES i

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._. . . - . . = _ - . . - -- -- _ - - - - _ _ . - _ - - - .-_-- -.

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i ITEM B.3:

LOW-POWER AND SHUTDOWN OPERATIONS RISK l

~

l (DR. POWERS) 4 l

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ITEM B.3: LOW-POWER AND SHUTDOWN OPERATIONS RISK l

The Committee provided a report to the Commission dated April 18,1997, regarding low-power and shutdown operations risk. The Committee has become concerned about the need for developing an understanding of the risk posed by the low-power and shutdown operations at nuclear power plants. This need arises because of repeated events during these modes of -

power plant operations, changes being made in plant operations in response to economic forces, and because of the ongoing initiative to develop risk-informed, performance-based regulation at the NRC. The staff will have to gain an understanding of risk during low-power and shutdown operations commensurate with its understanding of risk during power operations.

l During the December 1996 ACRS meeting, the Committee heard presentations by and held discussions with representatives of the Office for Analysis and Evaluation of Operational Data (AEOD) concerning the Human Performance Event Database. Information from this database indicated that more than 50 percent of the events investigated by Augmented Inspection Teams have occurred when plants were in low-power or shutdown modes. The evaluations of potential severe accidents during low-power and shutdown conditions at Surry and Grand Gulf show that low-power and shutdown risk is a significant factor compared to the risk calculated for the same plants during power operations. Based on previous evaluations of low-power and shutdown operations risk, and the above facts, the Committee provided the April 18,1997 repon to the Commission. This report contains several comments and recommendations, including the following:

• It is essential for the success of the Commission's effons to adopt risk-informed, performance-based regulation that a more complete understanding of the full spectmm of risk be established on a defensible technical basis.

• At present, there is no defensible regulatory basis to determine the extent to which results of risk analyses for power operations ought to be augmented to account for risk of low-power and shutdown operations. A more complete understanding of risks

~during all phases of nuclear plant operations is essential to ensure that regulations address real, significant risks and do not impose ad hos measures to correct discovered deficiencies in the hope that these measures will also address risk-significant issues.

• Significant effons may be needed to establish new risk assessment methods and to ,

understand phenomena associated with core damage events and the dispersal of radioactivity during low-power and shutdown operations.

• The NRC staff needs to establish a high-quality benchmark on risk during low-power and shutdown operations comparable to that which it has derived for risk during power operations.

l 51 9

., - - - , - m-- -

• - A well-planned, deliberate effon with realistic time schedules and extensive peer review should be undertaken first to develop methods and technologies that may be needed j- and then to benchmark risk during low-power and shutdown operations.

The Committee also provided a letter to the Executive Director for Operations on June 4, 1996, which was based on the Committee's review of the proposed rule on shutdown operations and of the shutdown risk studies performed for the Surry and Grand Gulf plants. ,

l In this letter the Committee made several comments, including the following:

• The concern for risk associated with shutdown operations has arisen from incidents l that have occurred. The quantitative understanding of risk posed by plants in low- {

l power or shutdown modes of operation is limited. l

• The available evidence suggests that shutdown operations can make important l l

contributions to the overall risk to the public posed by nuclear power plants. There l are no complete, reliable assessments of risk during shutdown operations even for a few  ;

representative plants. Cenainly, there is nothing commensurate with the NUREG-i 1150 study of risk during full-power operation.

i l

• The staff effon toward an interim solution by promulgating the proposed shutdown l operations rule is based on engineering judgment and will probably lessen risk. A risk- l informed understanding will require a quantitative evaluation of risk during low-power and shutdown operations. Therefore, priority attention should be given to performing l Level 3 PRAs for shutdown operations for the NUREG-1150 plants with consideration l l of spent fuel pool risk and uncenainty assessments.

f-

! The Committee plans to review the shutdown operations mle, which will also address spent fuel pool operations. The Committee also plans to review the extent to which low-power and shutdown operations risk will be incorporated into the development of risk-informed and performance-based regulations and regulatory guidance.

l l'

Attachments:

• Repon dated April 18,1997, from Robert L. Scale, ACRS Chairman, to Shirley Ann Jackson, NRC Chairman,

Subject:

Establishing a Benchmark on Risk During Low-Power and Shutdown Operations (pp. 54-59) j

• Letter dated June 4,1996, from T. S. Kress, ACRS Chairman, to James M. Taylor, Executive Director for Operations, NRC,

Subject:

Proposed Rule on Shutdown l Operations (pp. 60-61)  ;

l l 52 i

4

- - - - . - - - _ . _ _ . _ _ _ _ e--- . m- , , , , , - - -

. _. . - - _ - .. . . . . - . . - ~ _ . .- - -..-, . .-. _ _ - . - - - .

r Letter dated June 28,1996, from James M. Taylor, Executive Director for Operations, NRC, to T. S. Kress, ACRS Chairman,

Subject:

Comments on the Proposed Rule on Shutdown Operations (p. 62) t l

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.# o UNITED STATES g l 8 o NUCLEAR REGULATORY COMMISSION 1 l 3 ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

% CASHIN 2 TON, D C. 20655

' e ,,, e *#g April 18, 1997 l

l The Honorable Shirley Ann Jackson l Chairman l U.S. Nuclear Regulatory Commission l Washington, D.C. 20555-0001

SUBJECT:

ESTABLISHING A BENCHMARK ON RISK DURING LOW-POWER AND l SHUTDOWN OPERATIONS This report is to draw attention to the critical need for developing an understanding of risk posed by low-power and shutdown operations at nuclear power plants. This need is apparent as a l result of: (1) repeated events during these modes of power plant operations, (2) changes being made in plant operations in response to economic forces, and (3) the ongoing NRC initiative to develop risk-informed, performance-based regulation. We believe it is

,l essential that the NRC staff undertake a quantitative examination of risk during low-power and shutdown operations at representative i nuclear power plants. That is, the NRC staff needs to establish a i high-quality benchmark on risk during low-power and shutdown I operations comparable to that which it has derived for risk during l

power operations from the NUREG-1150 study (Ref. 1] and other sources. The benchmark for risk during low-power and shutdown l operations should address the following:

e a representative range of plant types, e all phases of low-power and shutdown operations, e accidents initiated by internal fires and other external events, e human performance, the unusual source term, radionuclide dispersal, and on-site populations that will affect the predictions of accident consequences, and l e uncertainties to a depth similar to that done for the risk '

benchmark for power operations.

l A substantial effort will be required to develop the technical l capabilities to conduct this benchmark risk analysis. Results of

( the benchmark risk analysis may suggest the need for refinements to the Commission's Safety Goals. In particular, the Commission may find from the results that it wants to specify limits on the

tolerable durations of plant configurations that pose very high risks.

54 ,

l

2 Our recommendation for a detailed benchmark analysis is based on l

the results of scoping risk studies done by the staff contractors i l [Refs. 2,3), the continuing string of worrisome events at plants ,

l during low-power and shutdown operations, and assessments of the risk significance of plant events by the Office for Analysis and  !

Evaluation of Operational Pata. The retaff's contractors have done l limited analyses of risk during one phase of shutdown operations at  ;

l a pressurized water reactor with a subatmospheric containment [Ref.

2) and one phase of shutdown operations at a Mark III boiling water reactor [Ref. 3). Results of these studies show that even when the l risk for a short period of shutdown operations is normalized over a full calendar year, the risk is a significant fraction of the risk calculated for the same plant during power operations:

Boiline Water Reactor Power Operations Shutdown Operations

• Mean Core Damage 4.1x10 2.1x10

Frequency Mean Early Fatality 8.2x10 4 1.4x10'8 Risk Mean Latent Cancer 9.5x10 3.8x10'3 Fatality Risk .

• Plant Operating Mode 5 (cold shutdown) only.

Pressurited Water Reactor Power Operations Shutdown Operations

• Mean Core Damage 4.1x10 5 4.2x10~6 Frequency Mean Early Fatality 2.0x10 4.9x10'8 l Risk Mean Latent Cancer 5.2x10*8 1. 6x10 2 l Fatality Risk l

I l

• Mid-loop operation only.

These partial results, however, may not adequately reflect current operating practices. The industry has instituted new guidelines

[Ref . 4) f or low-power and shutdown operations that are intended to i reduce risk. Several licensees are using software such as the Electric Power Research Institute's ORAM (Outage Risk Assessment 4

55

l 3

and Management) to plan activities during low-power and shutdown operations. These software tools are based on risk insights derived from simplified probabilistic risk assessments. If the NRC staff is to provide effective safety oversight of low-power and shutdown operations, the staff will have to understand the technical bases of the software tecls and the approximations in risk assessments that have been used to develop these tools. The availability of benchmark risk assessments for low-power and shutdown operations for representative plants akin to the benchmark risk assessments for power operations appears to be essential for the development of this understanding.

Despite the new guidelines and software tools developed by the l

nuclear power industry for low-power and shutdown operations,

! events that reveal safety vulnerabilities of the plants continue to l occur. Among the more recent of these events are:

• The Wolf Creek plant was in a " hot shutdown" condition when

activities involving the residual heat removal system created a flow path that allowed approximately 9,200 gallons of

, reactor coolant to transfer to the refueling water storage tank. Had this draining not been promptly terminated, the operability of the emergency core cooling system would have been compromised. The Accident Sequence Precursor Analysis indicated that this event had a high conditional core damage probability [Ref. 5]. The scoping studies of shutdown risk, however,. suggested that a " hot shutdown" condition was of such a low risk significance that it did not merit quantification.

e Loss of core cooling was threatened by the formation of a nitrogen bubble in the reactor coolant system at the Haddam Neck plant as a result of an improper valve lineup. Injection of high-pressure nitrogen into the reactor vessel continued for over three days while the plant was in a " cold shutdown" condition. The water'1evel in the reactor vessel was believed to have been displaced three feet below the vessel flange (Ref. 6].

l

• At the Cooper plant, about 10,000 gallons of water was inadvertently lost from the refueling cavity because a i submerged valve was opened to the main steam line drains. It I took over an hour for operators to identify the source of the loss of coolant inventory (Ref. 7].

Events during low-power and shutdown operations are consuming significant staff resources. At our meeting on December 5, 199., 6 we were told that more than 50 percent of recent events requiring Augmented Inspection Teams have occurred when plants were in low-power or shutdown conditions. Human errors during these conditions appear to be especially probable. A number of incidents that have l

56

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occurred during low-power and shutdown conditions are reviewed in the report NUREG/CR-6093 (Ref. 8). This report concluded that factors influencing operator actions are different from those typically regarded us important during full-power operations and states that: "Unlike full-power operations, large numbers of multiple concurrent tasks are possible during LP&S (low-power and chutdown) conditions. This has implications for both the ^PRA (probabilistic risk assessment) modeling process and the HRA (human l reliability assessment) quantification process."

We are concerned that this situation will be exacerbated as the industry moves to longer cycle times with less frequent l opportunities to exercise its low-power and shutdown operating procedures. The situation may also be exacerbated by industry efforts to shorten the duration of low-power and shutdown operations by increasing the intensity of activities during these periods. The industry will want to relieve burdens during outages -

by doing some maintenance while the plant is operating. For the staff to approve a trade-off between maintenance "on-line" and maintenance during outages, it will have to consider risk. To do l this, the staff will have to gain an understanding of risk during low-power and shutdown operations commensurate with its j understanding of risk during power operations.

The staff is now embarked on an effort to develop risk-acceptance l criteria for providing regulatory relief to licensees. Staff judgments on these matters Mre based on a firm foundation concerning event probabilities curing power operations derived both from the Individual Plant Examinations done by licensees and from its own benchmarking risk studies reported in NUREG-1150. There is l no comparable basis for making judgments concerning the accident probabilities and risk during low-power and shutdown operations.

At present, there is no defensible regulatory basis to determine the extent to which results obtained for power operations ought to be augmented to account for risk of low-power and shutdown operations. A more complete understanding of risks during all phases of nuclear plant operations is essential to ensure that regulations address real, significant risks and do not impose ad h2g measures to correct discovered deficiencies in the hope that these measures will also address risk-significant issues.

We believe it is essential for the success of the Commission's offort to adopt risk-informed, performance-based regulation that a more complete understanding of the full spectrum of risk be ostablished on a defensible technical basis. This more complete understanding is needed now as pivotal decisions are being made on the implementation of risk-informed, performance-based regulation.

We do not believe that existing scoping analyses or further scoping efforts will establish adequate benchmarks concerning risk during low-power and shutdown operations. This is especially so in light I

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5 is important.

time-dependent human performance of evidenceefforts Significant that may be needed to establish new risk assessment methods and to understand phenomena associated with core damage events and the dispersal of radioactivity during these phases of l plant operations. We are confident that areas of substantial

[ uncertainty will arise in the assessment Defensibleofquantification risk dur*ng low-power of these j

and shutdown operations.

uncertainties will require the same type of effort that was needed to quantify uncertainties in risk during power operations.

It will take time to develop a usefully complete understanding We recommend that of risk during low-power and shutdown operations.

a well-planned, deliberate effort with realistic time schedules and extensive peer review be undertaken first to develop methods and technologies that may be needed and then to benchmark risk during low-power and shutdown operations.

Sincerely yours, or R. L. Seale Chairman NUREG-1150, Vol. 1,

References:

Commission,

1. U.S. Nuclear RegulatoryAn Assessment for Five U.S. Nuclear

" Severe Accident Risks:

Power Plants," December 1990.

2. U.S. Nuclear Regulatory Commission, NUREG/CR-6144, Potential Severe Accidents During Low Power Summary and Shutdown of Results, October 1,"

Operations at Surry, Unit 1995. NUREG/CR-6143, SAND 93-U.S. Nuclear Regulatory Commission, 3.

2440, Vol.1, Sandia National Laboratories, " Evaluation of Severe Accidents During Low Powerofand Summary Results, Shutdown July Potential 1,"

operations at Grand Gulf, Unit 1995.

NUMARC 91-06, Nuclear Management andtoResources Assess Council, Shutdown Inc. ,

4. Industry Act. ions

" Guidelines for Management," December 1991.

5.

V S. Nuclear Regulatory Commission, in a Shutdown Information Noti Less of Emergency Mitigaticn Mctions While 1996, condition," dated March 25, 6.

U.S. Nuclear Regulatory Commission. Irformation Notice 94-36, Supplement 1, " Undetected Accumulation of Gas in Reactor Coolant System," November 1996.

58

6

7. U.S. Nuclear Regulatory Commission, AEOD/S96-02, " Assessment of Spent Fuel Pool Cooling," September 1996.

8. U.S. Nuclear Regulatory Commission, NUREG/CR-6093, "An  :

Analysis of Operational Experience During Low Power and Shutdown and a Plan for Addressing Human Reliability Assessment Issues," June 1994.

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UNITED STATES 8 n NUCLEAR REGULATORY COMMISSION

$ 0 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g

/[ WASHINGTON,,0. C. 20665 June 4, 1996 I i

Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

PROPOSED RULE ON SHUTDOWN OPERATIONS During the 431st meeting of the Advisory Committee on Reactor l Safeguards, May 23-25, 1996, we held discussions with l representatives of the NRC staff and the Nuclear Energy Institute l (NEI), concerning the subject proposed rule and the probabilistic j risk assessment (PRA) studies that were performed for the Surry and '

the Grand Gulf Nuclear Power plants. Our Subcommittee on Plant Operations met with the staff, NEI, and a utility representative on May 21, 1996, to discuss these matters. We also had the benefit of the documents referenced.

We previously commented on the staff  !

effort to resolve the shutdown operations issue in our letters dated August 13, 1991, April 9, 1992, September 15, 1992, and May 13, 1994.

According to the staff, the proposed rule will contain performance-l based elements. Since the supporting regulatory analysis and j

regulatory guide are still being developed, we discussed only the i proposed rule during our meeting. The staff has held sever'.1 public meetings with NEI to obtain industry input on the

formulation of this rule.

t

)

4 We made a number of comments on the risk basis for the rule. The staff agreed to consider our comments as it finalizes the draft rule, which it plans to publish for public comment in september l 1996. We plan to provide comments on the proposed final rule af ter <

] the staff has reconciled the public comments.

1

! The concern for risk associated with shutdown operations has arisen

from incidents that have occurred. Our quantitative understanding

of the risk posed by plants in low-power or shutdown modes of i

operation is limited. Risk assessments for shutdown operations were performed for Surry (a three-loop PWR with loop isolation

valves and a sub-atmospheric prersure containment) and Grand Gulf 1

(a BWR-6 with a Mark III containment). Neither of these plants is

) a particularly good surrogate for the entire population of PWRs and BWRs.

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The studies of shutdown risk consisted of two phases. The first phase was a deliberately conservative scoping analysis. The second phase focused on a single, high-risk plant operational state among the many that exist during shutdown operation. Such an approach could lead to an incorrect assessment of risk (a historical analogue is the selection of the large-break, loss-of-coolant accident as a bounding event) or to the adoption of operating practices that might increase risk.

The available evidence does suggest that shutdown operations can make important contributions to the overall risk to the public

. posed by nuclear power plants. On the eve of our entry into an era of risk-informed rulemaking, there are no complete, reliable assessments of risk during shutdown operations even for a few representative plants. Certainly, there is nothing commensurate with the NUREG-1150 study of risk during full-power operation.

The staff effort toward an interim solution by promulgating this <

proposed rule is based on engineering judgment and will probably lessen risk. A risk-informed understanding will require a quantitative evaluation of risk during low-power and shutdown operations. We therefore recommend that priority attention be given to performing Level 3 FRAs for shutdown operations at the NUREG-1150 plants with consideration of spent fuel pool risk and uncertainty assessments.

Sincerely, J 3. /@

T. S. xtess Chairman

References:

1. Memorandum dated April 5, 1996, from Robert C. Jones, Office of Nuclear Reactor Regulation, to John T. Larkins, ACRS,

Subject:

Development of $50.67, " Shutdown Operation of Nuclear Power Plants"

2. U. S. Nuclear Regulatory Commission, Prepared by Brookhaven National Laboratory, NUREG/CR-6144, " Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1," Summary of Results, October 1995

3. U. S. Nuclear Regulatory Commission, Prepared by Sandia National Laboratories, NUREG/CR-6143, " Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Grand Gulf, Unit 1," Summary of Results, July 1995

4. Nuclear Management and Resources Council, Inc. , NUMARC 91-06,

" Guidelines for Industry Actions to Assess Shutdown Management," December 1991 61

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• UNITED STATES g NUCLEAR REGULATORY COMMISSION d

WASNINGTON, D.C. SAAAM401 l June 28, D96

\ *M.# Thomas S. Kress, Chairman -

Advisory Comittee on Reactor Safeguards U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001 i

i

SUBJECT:

YOUR COMENTS ON THE PROPOSED RULE ON SHUTDOWN OPERATIONS (TAC N0. N77701) i

Dear Dr. Kress:

Thank you for the comments in your letter of June 4,1996, on the proposed rule 650.67, " Shutdown Operation of Nuclear Power Plants." We agree that available evidence does suggest that shutdown operations can be important contributors to the overall risk to the public. We also recognir.4 that the risk assessments for the Surry and the Grand Gulf plants canM be used as j

surrogates for the entire population of pressurized water reactors and boiling 4

water reactors. Nonetheless, we are using those insights inY,ch are generic to supplement our deterministic assessments in performing the regulatory

! analysis.

You noted that the proposed rule is based significantly on engineering

judgment. While this is true, this judgment draws upon many years of i

i experience with shutdown operations, a good understanding of events that have i

occurred and the causes of those events, and accounts for those shutdown activities which the staff's shutdown study,. documented in NUREG-1449, found

! to be most important. In addition, as part of our regulatory analysis, we are i

performing a probabilistic risk assessment (PRA) focused on those initiating l events which were judged to be most risk significant.

1 We recognize that following completion of our regulatory analysis, our quantitative understanding of shutdown risk will still be incomplete, and will not be comparable to the understanding gained from NUREG-1150 for power i

operation. You recommended that the staff perform a Level 3 shutdown PRA for the NUREG-1150 plants. While such a study may be desirable, we do not believe it is necessary in light of our ongoing regulatory analysis efforts.-

Furthermore, individual plant configurations during shutdown operation are more variable than at power. This not only complicates the performance of the risk study, it generally limits the generic applicability of the results.

i Thetefore, we believe that a full Level 3 PRA of the NUREG-1150 plants is not l warranted.

5

Sincerely,

-_/

ylor ecutiv Director for Operations i ec: Chairman Jackson

, Comissioner Rogers

Comissioner Dicus 62

! SECY a

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l ITEM B.4:

STATUS OF ACRS REVIEW OF THE NATIONAL ACADEMY OF SCIENCES / NATIONAL RESEARCH COUNCIL PHASE 2 STUDY REPORT ON DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS (DR. MILLER) l l

8

t ITEM B.4: STATUS OF ACRS REVIEW OF THE NATIONAL ACADEMY OF SCIENCES / NATIONAL RESEARCH COUNCIL PHASE 2 STUDY REPORT ON DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS During the December 6,1996 meeting with the Commissioners, the ACRS discussed the issues associated with the use of digital instmmentation and control (I&C) systems in nuclear power plants. The ACRS discussed the status of its review of the proposed update to Standard Review Plan (SRP), Chapter 7, " Instrumentation and Controls," Branch Technical Positions (BTPs), and regulatory guides for digital I&C systems. The ACRS also discussed the results of its review of the National Academf of Sciences / National Research Council (NAS/NRC)

Phase 1 study repon.

On October 13,1995, the ACRS issued its repon to the Commission regarding NAS/NRC Phase 1 study. In this report, the ACRS stated that the issues identified in the NAS/NRC  ;

Phase 1 study report will be imponant considerations as digital technology is used more l I

extensively in nuclear power plants. The Phase 1 study report identified the following eight key issues - six technical and two strategic:

Technical

• software quality assurance

• common-mode software failure potential

• system aspects of digital I&C technology

• human factors and human-machine interfaces

• safety and reliability assessment methods

• dedication of commercial off-the-shelf hardware and software Strategic

• case-by-case licensing

• adequacy of technical infrastructure The NAS/NRC Phase 2 study report was completed on January 31,1997. During its meeting of March 6-8,1997, the ACRS discussed the Phase 2 study report with representatives of the NAS/NRC Committee and the Nuclear Regulatory Commission staff. In pedorming its Phase 2 study, the NAS/NRC Committee limited its work to those issues identified in the Phase 1 study. The specific charge for the Phase 2 study was as follows:

• Identify criteria for review and acceptance of digital I&C technology in both retrofitted reactors and new reac: ors of advanced designs.

!

• Characterize and evaluate alternative approaches to the cenification or licensing of the l digital I&C technology.

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• Where sufficient scientific basis exists, recommend guidelines on the basis of which the Nuclear Regulatory Commission can regulate and cenify (or license) digital I&C technology, including means for identifying and addressing new issues that may result from future development of this technology.

• Where sufficient scientific basis exists to make recommendations, suggest ways in which i the Nuclear Regulatory Commission could acquire the required information.

i The NAS/NRC Committee recommends that both the nuclear industry and the Nuclear i Regulatory Commission be more proactive in participating in the relevant technical l communities and strengthen infrastructure in digital IRC systems. The communication barriers among panicipants could be addressed systematically. The NAS/NRC Committee also noted that licensing criteria should be forged in a detailed interaction among the l regulators, the industry, and the public. Generally, the NAS/NRC Committee emphasized I the following: l 1

= Deterministic assessment methods, including design basis accident analysis, hazard analysis, and other formal analysis procedures are applicable to digital systems, as long l as they are applied with care.

• Software failure probability can be used for the purporn of performing PRA to determine the relative influence of digital system failun. ou the overall system.

Including software failures in PRA is preferable to the alternative of ignoring software failures.

• Hardware and software must be treated together as a system; focusing solely on one or the other should be done with great caution.

• Most practical I&C systems cannot be exhaustively tested and thereby shown to be error free. However, adequate approaches exist and can be applied within practical resource constraints to support using digital systems in safety-critical applications in nuclear power plants.

The staff is in the process of incorporating the insights from the NAS/NRC Phase 2 study into the proposed final SRP, BTPs, and regulatory guides associated with digital I&C systems. The ACRS decided to postpone its comments on the NAS/NRC Phase 2 study report until after a meeting of its Subcommittee on I&C Systems and Computers, to be held on May 28-29,1997, during which the Subcommittee will discuss the views of the Nuclear Regulatory Commission staff on the recommendations included in the NAS/NRC Phase 2 study report and other issues related to the Standard Review Plan (Chapter 7 update).

1 1

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l Attachments:

i

• Report dated October 13, 1995, from T. S. Kress, ACRS Chairman, to Shirley Ann .

Jackson, NRC Chairman,

Subject:

National Academy of Sciences / National Research Council Study on " Digital Instrumentation and Control Systems in Nuclear Power l i

l Plants, Safety and Reliability Issues" - Phase 1 (pp. 66-67)

• 12tter dated October 31, 1995, frem James M. Taylor, Executive Director for  ;

Operations, to T. S. Kress, ACRS Chairman,

Subject:

"The National Academy of i Sciences' Report on Digital Instrumentation and Control, Safety and Reliability Issues" (p. 68) 1 1

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[ 9,, UNITED STATES i

l NUCLEAR REGULATCRY COMMISSION j $ ,' ADVISORY C0hANTTEE ON CEACTOR SAFEIUARDS 4

maanmaron.o.c. asses

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October 13, 1995 i

j

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~

1 The Honorable Shirley A. Jackson i Chairman l U.S. Nuclear Regulatory Commission j Washington, D.C. 20555-0001  !

j

Dear Chairman Jackson:

i j

SUBJECT:

NATIONAL ACADEMY OF SCIENCES / NATIONAL RESEARCE COUNCIL i STUDY ON " DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS IN 2

NUCLEAR POWER PLANTS, SAFETY AND RELIABILITY ISSUES" -

! PEASE 1 i

j During the 425th meeting of the Advisory Committee on Reactor l Safeguards, October 5-7, 1995, we reviewed the National Academy of

, Sciences / National Research Council (NAS/NRC) Phase 1 report on 1

2 Digital Instrumentation and Control Systems in Nuclear Power Plants, Safety and Reliability Issues. The NAS/NRC Committee i Chairman described the results of the Phase 1 report. We also had

the benefit of the documents referenced.

l The objective of the Phase 1 study was to define the important safety and reliability issues concerning hardware, software, and j human-machine interfaces that arise from the use of digital 4

instrumentation and control technology in nuclear power plant l operations. The report identifies eight key issues: six technical

{ and two strategic. It notes that these issues are common to other j industries where software is required for dependable operation of j systems. The report succinctly presents the issues that the l NAS/NRC Committee found to be important.

I We agree that the issues identified in the Phase 1 report will be

! important considerations as digital technology is used more

! extensively in nuclear power plants. In the past, we have called j attention to the effects of environmental stressors. The NAS/NRC i Chairman stated that the NAS/NRC Committee considered, but decided not to raise this issue to the level of a " key technical issue."

~

We continue to believe this is an important issue that the staff must address as it develops its regulatory guidance for digital 1 systems. However, this is part of the broader issue of i environmental qualification of safety-related equipment and does j not need to be a key issue of the Phase 2 study.

)i 66 i

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Honorable Shirley A. Jackson 2 1

j Me have concerns regarding a potential conflict between the Phase 2 completion schedule and the staff's schedule for issuing the Standard Review Plan-(SRP) and associated regulatory guides. Ne believe it is important that the SRP and other regulatory guidance 1 benefit from the insights in the Phase 2 report.

Sincerely,

h. 75 i T. S. Kress j Chairman i

4

References:

1 1. Report dated 1995, from the Cosmittee on Application of i Digital Instrumentation and Control Systems to Nuclear Power i Plant operations and Safety, Board on Energy and Environmental l Systems, Commission on Engineering and Technical Systems, j National Research Council, subject: Digital Instrumentation and Control Systems in Nuclear Power Plants, Safety and Reliability Issues - Phase 1

2. Memorandum dated December 2,1993, from Ivan Selin, Chairman,

! NRC, to NRC Commissioners,

Subject:

Computers in Nuclear l Power Plant Operations

! 3. Letter dated July 14, 1994, from T. S. Kress, Chairman, ACRS,

} to Ivan Selin, Chairman, NRC,

Subject:

Proposed National i Academy of Sciences / National Research Council Study and Workshop on Digital Instrumentation and control Systems

4. Letter dated August 23, 1994, from Ivan Selin, Chairman, NRC,

, to T. S. Kress, Chairman, ACRS, regarding ACRS letter of July

14, 1994 on National Academy of Sciences / National Research

. Council Proposal for a Study and Workshop on the " Application

! of Digital Instrumentation and control Technology to. Nuclear j Power Plant Operations and Safety" k

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i 67 j

j[%

c y *, UNITED STATES

S NUCLEAR REGULATORY COMMISSION j was m orow,o.c. es aan

• % *****/ October 31, 1995 Dr. Thomas S. Kress, Chairman Advisory Comunittee on Reactor Safeguards U.S. Nuclear Regulatory Comunission Washington, D.C. 20555-0001

SUBJECT:

THE NATIONAL ACADENY 0F SCIENCES' REPORT ON DIGITAL INSTRUNENTATION Ale CONTROL, SAFETY AIS RELIABILITY ISSUES

Dear Dr. Kress:

I am responding to your letter on this subject, dated October 13, 1995, in which the Advisory Comunittee on Reactor Safeguards (ACRS) comumented on the National Academy of Sciences / National Research Council (NAS/NRC) Phase 1 Report, " Digital Instrumentation And Control Systems In Nuclear Power Plants, Safety And Reliability Issues."

We agree that the issue of environmental stressors is key to the qualification-of safety-related digital instrtamentation and control systems. Environmental stressors are defined as an issue in the NAS/NRC study Phase 1 report, but not as a " key technical issue." As you know from past briefings to the ACRS, the staff is conducting confirmatory research to investigate and characterize the failure modes and degradation mechanisms of digital technologies proposed for use in nuclear power plants. Furthermore, this research is assessing the impact of smoke on advanced instrumentation and control hardware in nuclear power plants. The goal of this research is to provide the technical basis for a regulatory guide on the environmental qualification of digital instrumentation and control systems. We informed the NAS/NRC Coansittee about our activities on this key issue in an October 17, 1995, meeting with them.

We share your concerns regarding a potential conflict between the Phase 2 completion schedule and the staff's schedule for issuing the Standard Review Plan (SRP) and associated regulatory guides. Our contract with NAS/NRC calls for the completion of the study by September 30, 1996, which includes the delivery of the Phase 2 report. The staff has expressed its concern and will continue to encourage a timely completion of the NAS/NRC study. We agree that it is important that the SRP and other regulatory guidance benefit from the ,

insights expected from the Phase 2 report.

Sincerely,

/

s M. T lor ecutive Director for Operations cc: Chairman Jackson Coassissioner Rogers 68 SECY

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l ITEM B.5:

HUMAN PERFORMANCE PROGRAM PLAN l

I (DR. APOSTOLAKIS) l 1

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I ITEM B.5: HUMAN PERFORMANCE PROGRAM PLAN

, The NRC staff established a Human Factors Coordination Committee (HFCC) in 1994, which

! was comprised of representatives from the Offices of NRR, RES, NMSS and the Regions. The central task of the HFCC was to develop a Human Performance Program Plan (HPPP), which was initially issued in August 1995 and subsequently updated in 1996. The ACRS was first briefed on the HPPP during its February 1996 meeting. Subsequent meetings of the Human Factors Subcommittee were held in September and December 1996, to review the details of I

this Plan.

During its February 1997 meeting, the Committee completed its review of the HPPP and provided a repon, dated February 13,1997, to the Commission. In this report the Committee made several comments and recommendations, including the following:

I

• The HPPP is not a plan. It is, instead, an inventory of human performance projects within the agency. The HPPP should state explicitly what its goals are, what research 1 efforts will be required to achieve these goals, and when and how it will be known that l they have been achieved. The ownership of the present plan is diffuse. The success of such a plan as well as its dynamic nature require that ownership of the entire plan be clearly assigned. l

• A well-planned research effort in human performance is urgently needed to support both the regulation of plant operations and the transition to risk-informed and performance-based regulation. The overall perspective that can be provided by high-level models of human performance would be helpful in the planning of this research effon. A number of such models are reviewed in NUREG/CR-6350, "A Technique for Human Error Analysis (ATHEANA)."

• The development of indicators of a good safety culture, the design of a meaningful human performance reponing system, and the impact of downsizing and deregulation on human performance should be major elements of the research effort. The human reliability analysis research project should also be part of the HPPP.

The Committee offered additional suggestions for improving the HPPP and expressed its intent to continue to work with the staff on this matter.

The Executive Director for Operations has responded to the above ACRS report in a letter dated April 10,1997. His responses to the above three points can be paraphrased as follows:

• Agree that a comprehensive Human Performance Program Plan needs to be developed.

The Office of Nuclear Regulatory Research will assume leadership for this effort. An agency-wide program Plan for human reliability assessment / performance evaluation will l be developed for review by the end of June 1997.

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• Agree that a comprehensive research plan is needed. Beyond providing a mechanism for coordinating current staff activities in these areas, the plan will articulate the conceptual relationships between human reliability analysis activities and those of human performance evaluation. In addition, the "ATHEANA" model, which was developed to improve human reliability analysis, will serve as the framework to guide activities associated with this initiative. Essential Plan elements will include: a mission l statement, definition of strategies for achieving the mission, and identification of l program areas to implement the strategies. ACRS will be briefed when the Plan is fully developed.

t

• Agree that research effort should include development of indicators of good safety l culture, identification of the impact of downsizing and deregulation on human j performance, and the design of a meaningful human performance reponing system.

As part of its planned program to conduct operating events analysis to suppon human performance evaluation and human reliability analysis (HRA), the staff expects to l

recommend reporting requirements to better support HRA and human performance evaluation and to modify the LER coding scheme to include more human performance information.

l The Committee plans to work with the staff in developing an effective Human Pedormance l Program Plan.

Attachments:

l

• Report dated February 13,1997, from R. L. Seale, ACRS Chairman, to Shirley Ann l Jackson, NRC Chairman,

Subject:

Human Performance Program Plan (pp. 71-73) i l

• Letter dated April 10,1997, from L. Joseph Callan, Executive Director for Operations, to R. L. Seale, ACRS Chairman,

Subject:

Human Performance Program Plan, with enclosure, Responses to ACRS' Conclusions and Recommendations (pp. 74-77) i I

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/ 4o UNITED STATES y o NUCLEAR REGULATORY COMMISSION g .I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20665 l

i l February 13, 1997 l

The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

HUMAN PERFORMANCE PROGRAM PIAN ,

During the ' 438th meeting of the Advisory Committee on Reactor Safeguards, February 6-8, 1997, we completed our review of the NRC activities identified in the Human Performance Program Plan (HPPP) .

Our Subcommittee on Human Factors met on September 20 and December 3, 1996, to review these activities. During these reviews, we had the benefit of discussions with representatives of the staff.

In your remarks of December 2, 1996, to all NRC employees, you stated:

As we move to an era of nuclear power industry restructuring and declining NRC and industry resources, it is imperative that we are able to diagnose potentially declining licensee performance as early as possible.

We agree with your assessment. We believe that an appropriate HPPP would contribute significantly to the development of such diagnostic tools.

Conclusions and Recommendations

1. The HPPP is not a plan. It is, instead, an inventory of human performance projects within the agency. The HPPP should state explicitly what its goals are, what research efforts will be required to achieve these goals, and when and how it will be known that they have been achieved. The ownership of the present plan is diffuse. The success of such a plan as well as its dynamic nature require that ownership of the entire plan be clearly assigned.

2. A well-planned research effort in human performance is i urgently needed to support both the regulation of plant and operations and the transition to risk-informed 71

. - - - - - . - - - . - . - - . - . . - - ~ - . . - . - - . - - . _ -

j i

1 j performance-based regulation. The overall perspective that l

, can be provided by high-level models of human performance {

j would be helpful in the planning of this research effort. A ,

3 number of such models are reviewed in NUREG/CR-6350. l I i

3. The development of indicators of a good safety culture, the I design of a meaningful human performance reporting systcm, and i

the impact of downsizing and deregulation on human performance l should be major elements of the research effort. l Discussion l 3

operational experience has shown that human performance is a major power plants.

factor in the safe operation of nuclear Understanding what can go wrong at a plant requires an integrated evaluation of both hardware and human performance; i.e., the plant i must be viewed. as a sociotechnical system. In particular, the term l

! " human error," which carries the implication that the operators are l to be blamed, is inaccurate in many instances and one must

investigate and understand the context within which plant personnel

) function. This context is determined by both the design and the

physical' conditions of the plant, as well as by the prevailing j- safety culture.

i

{ The development of a plan for research on human factors is

! certainly not a simple task. This task would be made easier and the recommendations more convincing if the task were guided by a

! high-level model that identifies the important elements that ,

influence the likelihood of unsafe human acts. Various models and i taxonomies have been proposed in the literature and some are i beginning to receive wide acceptance. Human performance models and i i

error classifications that could be suitable guides for developing a research plan are being used in other projects in the office of Nuclea~ Regulatory Research. The models discussed in l NUREG/ CR -0350, along with insights from operational experience, '

could serve to guide the development of an HPPP.

One specific element we would like to see addressed in the HPPP is i the impact of situational assessment on compliance with procedures.  ;

Investigations of actual incidents and simulator exercises from nuclear and other industries have demonstrated the importance of what Professor James Reason of the University of Manchester calls

" intended violations" (circumventions) of procedures by plant personnel. The researchers who collected data from simulator exercises point out that these were not necessarily errors; the operators simply did what they felt was the optimal response to the evolving accident. We believe there is a need to understand the reasons for such deviations and how training, procedures, and the ,

plant safety culture could be modified to eliminate

" circumventions" to the extent possible. i 72 i

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1 The present HPPP contains elements that are worth pursuing. Other i eierents that should be contained in the HPPP include activities to

! gain a better understanding of the concept of safety culture and to j develop indicators of a good safety culture. The human reliability

analysis resea ch project should also be part of the HPPP. We will

continue to work with the staff in developing an effective HPPP.

j Sincerely,

R. L. Seale j Chairman 4

i

References:

I 1. Memorandum dated July 31, 1996, from Cecil Thomas, Office of 3 Nuclear Reactor Regulation, to John Larkins, ACRS Executive i

Director,

Subject:

Forwarding Human Performance Plan Rev. 1 l 2. Office for Analysis and Evaluation of Operational Data Report i j E-95-01, '* Operating Events with Inappropriate Bypass or Defeat l

! of Engineered Safety Features," July 1995  ;

4

3. U. S. Nuclear Regulatory Commission, NUREG/CR-6093, "An J Analysis of Operational Experience During LP&S and a Plan for Addressing Human Reliability Issues," June 1994

4. U. S. Nuclear Regulatory Commission, NUREG/CR-6265, l i "Multidisciplinary Framework for Analyzing Errors of l Commission and Dependencies in Human Reliability Analysis," i i

August 1995

5. U. S. Nuclear Regulatory Commission, NUREG/CR-6350, "A 4 Technique for Human Error Analysis (ATHEANA)," May 1996

) 6. Reason, J.T., Human Error, Cambridge University Press, j Cambridge, United Kingdom, 1990 i 7. R. Montmayeul, F. Mosneron-Dupin, and M. Llory, "The

Managerial Dilemma between the Prescribed Task and the Real j Activity of Operators: Some Trends for Research on Human

} Factors," Reliability Enaineerina and System Safety, 45:67-73, 1

1994

! 8. U. S. Nuclear Regulatory Commission, NUREG/CR-6208, "An

Empirical Investigation of Operator Performance in Cognitively i Demanding Simulated Emergencies," July 1994

! 9. International Atomic Energy Agency, Vienna, International

! Nuclear Safety Advisory Group, " Safety Culture," Report 75-INSAG-4, 1991

} 10. NRC Chairman Shirley Ann Jackson's remarks to all NRC j employees, December 2, 1996 l

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I 73 i

en arg gp 1 UNITED STATES j

t NUCLEAR REEULATORY COMMISSION WASHINGTON, D.c. Sanaa =1 l

\*****/ April 10, 1997 l Dr. Robert L. Seale, Chairman Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

HUMAN PERFORMANCE PROGRAM PLAN ,

l This responds to your letter dated February 13,1997, in which you provided j comments concerning the Human Performance Program Plan. Your conclusions and recommendations and the staff's responses to each are enclosed.

1 Sincerely, l L. o ep Callan Ex utive Director for Operations

Enclosure:

As stated cc: Chairman Jackson Commissioner Rogers Commissioner Dieus Commissioner Diaz Commissioner McGaffigan SECY OGC t

l 74

Enclosure Resoonses to ACRS' Conclusions and Recommendations

1. The HPPP is not a plan. It is, instead, an inventory of human performance projects I within the agency. The ;1PPP shoulv state explicitly what its goals are, what research efforts will be required to achieve these goals, and when and how it will be known that they have been achieved. The ownership of the present plan is diffuse.

The success of such a plan as well as its dynamic nature require that ownership of the entire plan be clearly assigned.

The ACRS is quite correct. As stated in our February 4,1997, responses to ACRS questions, the HPPP was intended to function as a mechanism to coordinate human factors activities among the Agency's offices. However, we agree that a comprehensive program plan needs to be developed.

Presently, ownership for coordination for the HPPP is in NRR. Since future needed efforts are developmental and are expected to involve comfirmatory research, leadership will be shifted to the Office of Nuclear Regulatory Research. An agency-wide program plan for I human reliability assessment and human performance evaluation is expected to be developed for review by the end of June.

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l 2. A well-planned research effort in human performance is urgently needed to support  !

l both the regulation of plant operations and the transition to risk-informed and  ;

performance based regulation. The overall perspective that can be provided by high level models of human performance would be helpful in the planning of this  !

l research effort. A number of such models are reviewed in NUREG/CR 6350. '

We agree that a comprehensive research plan is neede<f. Consistent with the ACRS's suggestion, the staff is developing a plan for integrating activities in human reliability i

assessment and human performance evaluation. Beyond providing a mechanism for 1

( coordinating current staff activities in these areas, the plan will articulate the conceptual relationships between human reliability analysis activities and those of human performance evaluation. In addition, the "ATHEANA" model, which was developed to improve human reliability analysis, will serve as the framework to guide activities associated with this i mitiative.

l l

l The essential elements of the plan will be a statement of the mission of the human reliability assessment and human performance evaluation initiative in the Agency, a definition of strategies for achieving the mission, and the identification of program areas l

established to implement the strategies. Within each program area, summaries of ongoing l'

l and planned activities will describe their relevance and consistency with the defined i

j mission, strategies, and programs. Such an approach to planning is expected to better highlight any redundancy, gaps, and disproportionate emphasis in current and proposed l

staff efforts. We will brief the ACRS on the program plan after it is fully developed.

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3. The development of indicators of a good safety culture, the design of a meaningful ]

human performance reporting system, and the impact of downsizing and deregulation on human performance should be major elements of the research effort.

We agree that the research effort should include development of indicators of good safety .

culture, identification of the impact of downsizing and deregulation on human performance, t

and the design of a meaningful human performance reporting system.

As part of its planned program to develop the technical basis and guidance on management and organizational influences in human performance and plant risk, the influences of l

l

management practices and safety culture on human performance and human reliability will j

j be identified. Additionally, the impact of downsizing and deregulation on human performance and human reliability will be investigated. As part of efforts to improve the Senior Management Meeting process, the staff is identifying measures of economic stress that can be used to identify plants for increased safety monitoring.

i 4

Licensees are required to report data to the NRC on factors that influence operating events, including f actors that contribute to human performance during events. However, the i

human performance data collected are not always sufficiently detailed to provide the basis

for formulating research or regulatory programs. As part of its planned program to conduct

operating events analysis to support human performance evaluation and human reliability

$ analysis, the staff expects to recommend reporting requirements to better support HRA and l

1 human performance evaluation and to modify the LER coding scheme to include more 1

human performance information.

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ITEM B.6:

ACRS REPORT TO CONGRESS ON NUCLEAR SAFETY RESEARCH AND REGULATORY REFORM l (DR. POWERS) l I

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ITEM B.6: ACRS REPORT TO CONGRESS ON NUCLEAR SAFETY RESEARCH AND REGULATORY REFORM The Advisory Committee on Reactor Safeguards is required by Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209 to report to Congress each

-year on the Safety Research Program of the U.S. Nuclear Regulatory Commission. ,

On February 21,1997, the ACRS provided its 1996 report to the Congress. In this report, i

~

the ACRS expressed its views on the potential impact of a reduced Safety Research Program on regulatory reform and the ability to provide adequate safety oversight for a changing nuclear industry.  ;

The highlights of the ACRS repon include the ' allowing:

• Continued availability of unbiased safety research information will be essential as the )

NRC establishes itself as the leader in the national effort to reform the regulatory 3 process to focus on real risks, continued safety of operating nuclear power plants, and J the performance of licensees.

• The Safety Research Program has enabled the NRC to develop a probabilistic risk assessment method that can provide quantitative measures of the real risks of nuclear power.

• Based on information that has come from the Safety Research Program, operational experience, and the ability to quantify risk, the NRC has been able set forth safety goals that define how safe is safe enough.

• Understanding of risk has reached the point that it can be used to reformulate the regulatory structure to focus both licensee and regulatory attentions on what is significant to safety in a cost-effective way.

• The Safety Research Program has aided the NRC in the development of standards for ,

the use of risk assessment in the regulatory process.

• A risk-informed regulatory structure will be essential for the NRC to respond to changes taking place within the nuclear industry in response to economic pressures.  !'

These changes include modernization of instrumentation and control systems, downsizing workforces, and extending fuel lifetimes. Each of these changes mandates  ;

l evolution of NRC rules and regulations to assure continued, adequate protection of the  !

public health and safety.

]

• Funding for the Safety Research Program has been reduced to a level that may not

allow a cost-effective response by NRC to the new challenges.  !

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• The Safety Research Program will have to be sustained and even augmented if the NRC is to complete its transformation to risk-informed and pedormance-based j regulatory approach. The NRC effon to establish a risk-informed and performance-  !

based regulatory approach is an example for other regulatory agencies that should not l be allowed to fail.  !

• - Without the needed research suppon, the NRC may be forced to rely on historical, l conservative, costly regulations not necessarily focused on risks. l The ACRS plans to continue its review of various elements of the NRC Safety Research Program.

Attachment:

+

• Repon dated February 21,1997, from R. L Scale, ACRS Chairman, to the Honorable

Alben Gore, Jr., President of the United States Senate, and to the Honorable Newt  ;

Gingrich, Speaker of the United States House of Representatives,

Subject:

The l Advisory Committee on Reactor Safeguards Repon on Nuclear Safety Research and Regulatory Reform (pp. 80-86) l l

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4 4

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,- [* o' , UNITED STATES S ,,

NUCLEAR REGULATORY COMMISSION M I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS  !

I WASHING 7oN, D. C. 20555 l

) k,.... ,

j February 21, 1997 i

The Honorable Albert Gore, Jr.

President of the United States Senate Washington, D.C. 20510 )

l

Dear Mr. President:

I am pleased to transmit to the Congress the 1996 report of the Advisory Committee on Reactor Safeguards on the U. S. Nuclear

Regulatory Commission's Safety Reseaach Program. This report is l required by Sec. tion 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209.

I Sincerely, I

R. L. Seale Chairman

Enclosure:

U. S. Nuclear Regulatory Commission, "The Advisory Committee on Reactor Safeguards Report on Nuclear Safety Research and Regulatory Reform," dated February 1997 s

80

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, a cte uq'o, UNITED $TATES y ,

NUCLEAR REGULATORY COMMISSION L'

I .

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20655 oY t

i February 21, 1997 1

The Honorable Newt Gingrich speaker of the United States House of Representatives Washington, D.C. 20515

Dear Mr. Speaker:

I am pleased to transmit to the Congress the 1996 report of the Advisory Committee on Reactor Safeguards on the U. S. Nuclear Regulatory Commission's Safety Research Program. This report is required by Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209.

Sincerely, R. L. Seale Chairman

Enclosure:

U. S. Nuclear Regulatory Commission, "The Advisory Committee on Reactor Safeguards Report on Nuclear Safety Research and Regulatory Reform," dated February 1997 i

l 81

i A REPORT TO THE CONGRESS  !

OF THE UNITED STATES OF AMERICA  !

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i ON l

l NUCLEAR SAFETY RESEARCH AND REGULATORY REFORM I l l

l BY l

l THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS U.S. NUCLEAR REGULATORY COMMISSION l

l FEBRUARY 1997 f

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82

s e l

l THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS REPORT ON NUCLEAR SAFETY RESEARCH AND REGULATORY REFORM )

l The Advisory Committee on Reactor Safeguards, in the past, reported

! on very specific reactor safety research issues and programs. In I light of the diminished resources available to support the U.S.

Nuclear Regulatory Commission's Safety Research Program, we have chosen, instead, to report on the potential effects of a reduced i Safety Research Program on regulatory reform and the ability to provide adequate safety oversight for a changing nuclear industry.

A vigorous research program dealing with the safety of commercial nuclear power production has served the Nuclear Regulatory Commission ano the public well in the past. The continued availability of unbiased safety research information will be essential as t're Nuclear Regulatory Commission establishes itself as the leader in the national effort to reform the regulatory process to focus on real risks, continued safety of operating nuclear power plants, and the performance of licensees. At the

same time, initiatives taken by the commercial nuclear power industry in response to ongoing and anticipated deregulation of electrical power generation make it even more important that the  !

Nuclear Regulatory Commission continue to have a Safety Research Program that provides the information needed to modify and improve its regulations to protect public health and safety.

From the inception of the civilian use of nuclear energy to generate electrical power, public safety has been of paramount concern. Initially, litt:.e experience and few industrial safety standards were available to ensure that nuclear power could be t generated safely. As a result, prescriptive, highly conservative approaches that blanketed all aspects of nuclear power generation 83

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+ .

2 were adopted by both the regulatory authority and the industry.

Faults and vulnerabilities identified through operation of nuclear power plants were used to add layers of protection on this ,

regulatory structure. Indeed, regulation of nuclear power generation has been successful in protecting public safety in this country. But, safety has been achieved through highly conservative regulation at great cost to both the producers and consumers of nuclear power.

As nuclear power generation has natured, experience has been gained in our understanding of the real risks of nuclear power. The Safety Research Program has enabled the Nuclear Regulatory Commission to develop a method called probabilistic risk assessment that can provide quantitative measures of these risks. The sophistication of this understanding has reached the point that it is now possible to initiate a reformation of the regulatory structure for nuclear power generation. This reformation will focus attention on what is significant to safety and at the same time will allow the industry to identify and use cost. effective ,

strategies to mitigate risks. Reformation of regulation of all types to focus on risk is, of course, a national priority. The Nuclear Regulatory Commission is taking the lead in this national effort with its policy of risk-informed and performance-based regulation. Based on information that has come from the Safety Research Program, operational experience, and the ability to quantify risk, the Nuclear Regulatory Commission has been able to set forth safety goals that define how safe is safe enough. By working with individuals experienced in plant operations and using the tools of risk analysis,the NRC can now identify regulations that do not contribute to safety, and it will be able to define a rational, cost-benefit basis for imposing additional regulatory requirements.

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i-Steps are being taken in the direction of risk-informed and '

l performance-based regulation. The performance-based maintenance j rule (10 CFR 50. 65) is a tangible accomplishment.

4 Rather than i imposing bureaucratic prescriptions on every aspect of safety

] system maintenance, this rule allows the industry to find creative j strategies to meet performance objectives approved by the Nuclear Regulatory Commission based on risk information. Satisfactory

] performance by licensees is rewarded by reductions in regulatory

] burdens while performance failures elicit increased regulatory l scrutiny.

l The Safety Research Program has aided the Nuclear Regulatory l Commission in the development of standards for regulatory use of 1

l risk assessment. This would permit additional uses of this

'a approach to focus dwindling resources on issues of most importance for protecting public health and safety. Target applications of j these new standards are in-service inspection, in-service testing,

,and technical specifications for reactor safety systems. Continued research will be essential for further regulatory reforms.

New challenges to the regulation of nuclear power are emerging.

These challenges come from the deregulation of electrical energy production and the need for the nuclear power industry to become more cost competitive. The nuclear industry is aggressively pursuing changes to remain economically viable. These changes could have significant safety implications that will require regulatory approval when they affect the licensing ' basis for l nuclear power plants. Among the changes under consideration are increased fuel lifetimes, elevated operating power, digital instrumentation and control systems, and downsized work forces.

Each of these changes could challenge the existing regulations for the protection of public health and safety. We believe that applied regulatory research programs will be required to develop bases / criteria for regulatory approval of these changes. of 85

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4 particular importance are the changes that may affect human performance in the operation of nuclear power plants.

P.nding for research activities has fallen by a factor of about 3 over the last 10 years and all evidence points toward continued reductions in the future. While much of this decrease can be attributed to the maturation of the technology, funding for the j Safety Research Program has been reduced to a level that may not i allow a cost-effective response to these new challenges. The Nuclear Regulatory Commission now does not have the technical tools needed to evaluate all of the safety implications of extending fuel '

lifetimes to the extent the nuclear industry has requested. It cannot evaluate quantitatively the risk implications of personnel reductic.ns and modernization that are being proposed by the nuclear l industry. The Safety Research Program will have to be sustained and even. augmented if the Nuclear Regulatory Commission is to complete its transformation to risk-informed and performance-based regulatory approach. Without the needed research support, the l Nuclear Regulatory Commission may be forced to rely on historical, conservative, costly regulations not necessarily focused on risks.

Safety innovations by the industry may be stifled. The opportunity to use regulation of nuclear power as an example of successful regulatory reform may be lost.

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