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Transcript of ACRS 303rd General Meeting on 850711 in Washington,Dc.Pp 1-165.Supporting Documentation Encl
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{{#Wiki_filter:P ORIGINAL UNITED STATES OF AMERICA ('~] C/ NUCLEAR REGULATORY COMMISSION In the matter of: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 303rd General Meeting Docket. No. /3U e e i I Location: Washington, D. C. Date: Thursday, July 11, 1985 Pages: 1-165 ) ACRSOFFICECOPY hk JreecAMSMice ourt leporters 1625 I St., N.W. 1 Suite 921 8507160284 850711 fashington, D.C. 30006 (202) 293-395Ct QQ12gcR5 q,

1 1 UNITED STATES OF AMERICA m 2 NUCLEAR REGULATORY COMMISSION 3 4 Advisory Committee on Reactor Safeguards 5 303rd General Meeting 6 7 1717 H Street, N.W. 8 Room 1046 9 Washington, D.C. 10 11 Thursday, July 11, 1985 12 13 The Advisory Committee on Reactor Safeguards met in 14 open session, pursuant to notice, commencing at 1:50 p.m., 15 David Ward, Chairman of the Committee, presiding. 16 ACRS MEMBERS PRESENT: 17 David Ward G. Reed 18 H. Lewis C. Wylie 19 J. Ebersole F. Remick 20 D. Moeller P. Shewmon 21 W. Kerr C Mark 22 M. Carbon C. Siess 23 D. Okrent R. Axtmann 24 H. Etherington Os 25

e 2 1 ALSO PRESENT: 2 R. Savio, ACRS Staff Member 3 E. Igne 4 H. Alderman 5 R. Fraley 6 PRESENTERS: 7 V. Stello T. Spies 8 T. Murley F. Rowsome 9 F. Gillespie S. Brocoum 10 S. Israel D. Brand 11 E. Jordan A. DeAgazio 12 B. Sheron G. Rivenbark 13 D. Powell H. Wong 14 15 16 17 18 l l 19 20 l 21 l l 22 23 24 l l 25 l

3 1 P RO C E ED I NGS 2 MR. WARD: Our next agenda item is No. V, 3 quantitative safety goals. I think we will start out with the 4 subcommittee report. Dr. Okrent, 5 MR. OKRENT: Well, Dick is handing out now a i 6 slightly revised version modified, or whatever you want to 7 call it, of the draft we were talking about on the Saturday of 8 the last meeting, and one can see what the changes are by 9 crossout and underlining. Okay? 10 Some of them are just editorial, some are for 11 purposes of discussion, some sort of seem like they might be 12 improvements as a result of things that have transpired ) (_/ 13 yesterday and today, et ceteta. 14 I should note that at the back of this, at the clip 15 you will see something in blue which is a draft of a possible 16 letter that I was asked to prepare, so that is something that 17 we would presumably look at Saturday. 18 Since it is not completely disconnected, in my 19 opinion, from some of the things that relate to safety goals, 20 and since I managed to.get an idea for a possible draft, it is 21 available now. 22 Other than that -- well, let's see 23 MR. WARD: Aren't we going to hear from the Staff? 7 24 MR. OKRENT: We are in a minute. There is an agenda 25 you have. Has that been handed out?

4 1 MR. WARD: Yes. j-ss 2 MR. OKRENT: Okay. And I would propose that, you 3 know, we had about a four-hour subcommittee meeting yesterday, 4 but rather than my trying to provide a summary of it now, 5 since we have most of the principal actors from yesterday here 6 today, I would propose unless someone else wants to make a 7 comment first, that we move into the part with the NRC Staff 8 now. 9 MR. WARD: Okay. Good. 10 MR. OKRENT: I assume we still have two hours with 11 this agenda item; is that correct? 12 MR. WARD: That's right. 13 MR. OKRENT: So in that case, Vic Stello is the 14 first one up. 15 MR. STELLO: Well, let me start by saying I had the 16 opportunity to sit and listen to the committee's discussion of 17 this item with the Commission, and I wanted to start by saying 18 it is somewhat troubling to me for two reasons: 19 The committee -- 20 MR. REMICK: I'm sorry, I'm having trouble hearing 21 you, Vic. 22 MR. STELLO: Well, it works. 23 The Committee continues to remind us that when there 4 24 are issues that they are particularly interested in, they want 25 to make sure that we make them aware that we are working on

5 1 those issues and give them copies of the appropriate papers 7s 2 and reports so they would have early input. 3 We did that with respect to the safety goals. We 4 clearly have not evolved to where we are prepared to say what 5 the Staff's view is on this issue. 6 We wanted to make sure that the committee was aware 7 of the variety of views on a number of issues that are 8 contentious. 9 Sitting listening to the discussions, it seems that 10 there were a number of comments that puts the Staff in what I 11 would characterise, I guess, fairly as a bad light, and_I 12 don't think that that's fair. O (,/ 13 If the committee wishes to heir what the Staff is 14 thinking before the Staff has a view,.then I think that's 15 healthy and I think it's proper and I think we ought to do 16 that. Then there has to Be an understanding that you are 17 hearing fairly rough ideas that have not had even a chance to 18 be adequately discussed within the Staff, and it is unfair to 19 over bring the result of.those discussions in any unfavorable 20 light. 21 It would be unfortunate if the net result is that 22 the committee -- in the way it came across at least to me -- 23 was suggesting that the only time they want to talk to the 24 Staff is when there is a final polished complete decision by 25 the Staff and the Staff is ready to stand behind that.

,.. =. .~ 6 1 I think that would be a big mistake. I think that .{b 2 the committee and the interaction of the committee with the 3 Staff in working out some of these very difficult areas is 4 very, very beneficial, and I quite frankly find it very, very and I have repeatedly said and I would start by 5 useful 6 suggesting again that there are a number of these areas that 7 if the committee cannot write a letter on this generic topic 8 of safety goals, there are at least a number of parts of it 9 which no matter what the safety goal finally says are 10 particularly difficulty issues, and perhaps the committee 11. could find a way in which to state its view regarding some of J 12 these difficult issues. 13 Those are the difficult issues, I guess, that I 14 wanted to go to next and would start by hoping that the 15 committee, if it is unable to write a definitive letter -- and 16 I understand there are a variety of views on the committee on at least we could focus on a number of 17 this subject 18 difficult subjects and get the committee's advice, counsel, if 19 it has to have half a dozen additional comments. I don't 20 think that that's bad, either, in trying to at least bring 21 into light some help on some of these issues. 22 The two that I emphasized when we were here last 23 time, the same two issues that I guess I want to emphasize 24 again today, to avoid any controversy, let me say that j 25 anything we say, we are going to mean "mean," whether it turns l

7 /"Ng 1 out that way or doesn't turn out that way. I don't think it's N,sY } 2 relative. I think a discussion of mean and median -- well, I ) l 3 just don't think it's very productive. 4 MR. LEWIS: If I could just interrupt you for a 5 second, I agree with that. And the raason for my frustration 6 -- and as we both know, I said the nastiest things this is that it has been around for at least two years, 7 morning 8 to my personal knowledge, probably longer than that, and so it 9 is hardly an issue which has just come up and on which the 10 Staff has not reached agreement. 11 I applaud what you just said. 12 MR. STELLO: Let me just suggest that it be in the ( 13 first paragraph you write in the letter. Get the committee to 14 agree that from now on whenever we talk about any of these 15 results, we mean "mean." Fine. If that's the committee's 16 view, say it. I think it would be useful. It may have made 17 that thing go away a year ago, I don't know. If you can't 18 deal with the whole of the subject, at least deal with the 19 parts of it that continue to be a problem and we can get them 20 out of the way. I don't think it's worth very much 21 discussion. 22 But there are two areas that I do believe are worth 23 a great deal of discussion, and that's the whole issue of () 24 where you set the performance criterion at. At 10 to the 25 minus, 10 to the minus 5. Do we mean by 10 to the minus 4 the

8 e 1 classical results of PRA which are predictors of the 2 capability for the core to be severely damaged, which may lead 3 to a full scale core melt, where the core physically melts 4 through the vessel, in contrast to a goal of 10 to the minus 5 5, for which the core is already calculated to have left the 6 vessel and its fission products then are inside the 7 containment. 8 I think that's a fundamental, philosophical point j 9 for which one has to go back and look at what can you get at, 10 as a result of the calculation? Can you ask the PRA I 11 technology to take that next step? 12 If we do, are we then really putting off using any s_/ 13 in any meaningful way safety goals until we have developed yet 14 even further PRA technology to be able to discriminate between 15 those two kinds of :alculations? 16 My preference is, I would rather we didn't do that, 17 we would use the PRA technology for what exists today and take 18 those numbers and use those numbers and just recognize that 19 there is some degree of conservatism in using them in that 20 manner, whether it's two or 10. I don't know that anybody 21 really knows. You've had numbers offered to you as to what 22 the differences are. 23 It would be very, very helpful if the committee can () 24 deal with that subject. That is a policy matter that the 25 Commission clearly has got to come to grips with, what ought

.-...-.=.-. -._.. i 9 1 that policy be, and we ought not to use any surrogates for how O 2 to get there. That is, we ought not to try to decide that a 'r 3 cost-benefit analysis doesn't produce the kinds of changes 4 we'd like to make and therefore let's make it some other 5 number. Let's just do it strictly on the matter of policy. 6 It is a policy issue as to where that ought to be 7 set. I think we ought to decide that issue and move on with 8 it. 9 Cost-benefit analysis. We spent a great deal of 10 time talking about cost-benefit analysis. There is a great 11 deal of controversy behind it. Where ought we go with this 12 issue? s 13 I was, I guess in terms of summarizing it -- we go 14 where we were before with the area of cost-benefit, which 15 would not include any of the averted costs. That's what the 16 draft safety goal was all about. Or we could go to the end of 17 the spectrum, which was suggested, as all costs, all benefits, 18 all of the time. 19 I think this again is a matter of policy as to 20 whether it's important to have that kind of information 21 displayed to decisionmakers, and the decisionmakers in this 22 agency may reside in a lot of places. This room -- you-all 23 are decisionmakers. Eventually you reach a conclusion and you 24 write a letter that says what it is that you have decided on g 25 an issue. _._ _ _,.._._ ___.i

l 10 4 1 The Director of NRR has to reach a conclusion or a 2 decision on a particular licensing matter. 3 Hearing Boards reach conclusions. 4 The EDO has to reach conclusions. The Commission. 5 What information ought that decisionmaker see with i 6 respect to these issues before he decides? I can say it as 7 simply as I know how: 1 8 I'd like to know everything that there is know about 9 the issue that's reasonably obtainable before I decide it. l 10 And I say that simply as saying personally I like to see all 11 costs, all benefits. 12 Saying it another way, I'd like to know everything F \\ 13 that I can find out about the issue before I decide it. f 14 Now, one other point that I don't know how to 15 emphasize enough, so it isn't misunderstood, as I gather there 16 is a lot of misunderstanding about: 1 17 There's nothing magic about safety goals. The 18 safety goal is not going to be a mechanism whereby we replace 19 the current regulatory process. It will be an additional 20 element in the regulatory process. The report from the 21 steering group, in fact, I think, used language something like 22 "we should not supplant the current traditional regulatory 23 methods that we use," the defense-in-depth concept. ( 24 I think that's stronger than we ought to say it. 25 But even in the extreme, it is clear that you are not going to

11 i use what you get from safety goals to change that regulatory (v-s h 2 process from the fundamental way we do things today. 3 If you look at the back-up report from the steering 4 committee, you will find that most of the decisions that have 5 been made that were tested with the safety goal would not have J 6 passed the safety goal standard of $1000 a manrnm. 7 Those requirements issued, even though they didn't 8 meet that cost-benefit balance. And I suspect that that's a 9 probably the kind of thing that will continue in the future. 4 10 But there ought to be a rationale. You ought to know what the 11 results of that cost-benefit analysis are. If you decide not 12 to go that way, there ought to be a reason for doing it k 13 On the other hand, if we get results of PRA that 14 show that our traditional methods of regulation suggest that 15 we're not doing something that's needed for safety, then by i 16 definition it ought to supplant those traditional methods and 17 we ought to require something where PRA methods show us we 18 ought to. If they clearly show us that we ought not to do 19 something, then too we ought to give that some substantial 20 weight. 21 I don't know that this is written out plain enough, 22 because it seems to create a great deal of confusion, at least 23 when I listen, as though there's some fundamental change ) 24 that's going to take place, should we ever adopt safety goals. 25 I don't think that that's advocated or suggested yet ~

12 1 by anyone. I think we are a long way from where we would make } d 2 a transition from our current regulatory process to a 3 completely new regulatory process derived from the kind of 4 methodology embodied.in the safety goals. We are not there. 5 But we're never going to learn a great deal, in my 6 view, until we get started. I prefer to start, and start 7 slow, do what we can, use it intelligently and rationally. If 8 it's being used improperly, I'm sure the ACES will remind us 9 of when the use is improper. We all ought to be watching each 10 other. But I would urge that we get on with it. I think it's 11 very, very important that we do that. 12 And again I want to plead that the ACRS try to pick at least these several difficult issues and try to 13 these 14 decide where it comes out on these difficult issues and at 15 least put down some sort of comments back so that we would 16 have the benefit of the counsel of the committee in these 17 areas. If it can't bring itself to deal with the whole issue, 18 I urge it at least deal with parts of the issue so we can make 19 as much progress as we can possibly make. 20 I think I've probably said more than enough 21 already. There are a number of other changes that I think 22 should have had more discussion when we gave them to the 23 implementation section, in particular, for which if you will ) 24 read that one very carefully, it says that that is to be 25 developed later by the EDO, and I don't think we ought to j

13 1 spend a great deal of time about it. f w., 2 If there is any particular issue about the and that's the one that describes how the 3 implementation 4 Staff would use results that are bounded from 10 to the minus 5 3 to 3 times 10 to the minus 5 in terms of results, and how we 6 would implement that. If we need to talk about that, then I 7 think we ought to. 8 We certainly didn't in the subcommittee. There are 9 a number of other changes that we talked about that I briefed 10 you about-last time that we're thinking about. 11 With that, I'd like to ask Themi Speis to summarize 12 the NRR position on this matter. V 13 [ Slide.) 14 MR. SPEIS: I am Themi Speis from the Office of NRR. 15 I am here to summarize some of the views that were 16 provided to the EDO by Harold Denton. Unfortunately he is on 17 vacation and decided to stay there. He is somewhere in North 18 Carolina. We did talk to him last night, and he was briefed 19 on what happened at the subcommittee meeting, and basically he 20 told me to relate to you people that he has been personally 21 involved in our safety goal deliberations, he has been briefed 22 by members of the steering group that came from the Office of 23 NRR. He has been briefed by Dr. Murley. He has been () 24 discussing this with Mr. Dircks, but he still is reviewing all 25 these important issues that you have been hearing and talking

14 I 1 about the last few days, and maybe the last few years. (1s) 2 So his views are still evolving, even though he is 3 not withdrawing the letter or anything like that. But, you 4 know, his work is input to the EDO and his final concurrence 5 will depend on the views that the other offices in EDO and 6 what views you might have and things of that sort. T 7 Frank Rowsome briefed most of you yesterday. I will 8 repeat some of the same things. 9 One of the things that Harold told me was to 10 separate the two viewgraphs that were presented yesterday and 11 make sure that in the first viewgraph I have the one serious 12' difficulty, so I put the other things he said about the safety C 13 goal in the other viewgraph. He doesn't want to mix the big 14 things with the small things. So I did that. It was very 15 easy to do it. 16 So you see that one of the difficulties that he has, 17 that he is suffering with this issue, is the goal itself, 18 whether it should be 10 to the minus 4 or 10 to the minus 5. 19 You are familiar with the 1 minus E to the minus-lambda d type 20 of calculations. If you assume you have 100 plants and if you 21 assume you have a coremelt probability of 3 times 10 to the 22 minus 4 and plot the numbers there for the next 20 years, 23 you'd come up with a 45 percent chance of having a serious () 24 accident in the next 20 years, and a 10 percent chance of two 25 or more such accidents. 4 _..___,__-_,.,.r,. _-_..,_._-r

15 s 1 So this kind of bothers him, you know. He thinks 2 that the number looks too big. 3 The other thing that bothars him is we can all say 4 that we can have a core melt accident but nothing will happen 5 because the containment is there and the appropriate systems 6 included within the containment will prevent any type of 7 radioactivity getting out, and even though he believes that 8 substantial progress has been made in this area, in both 9 analytical and experimental to some extent, and we know more 10 about what type of challenges are generated inside the 11 containment as a result of a serious accident and how 12 containments respond to the challenge, he still feels that 13 maybe we are putting too much reliance on those types of 14 analyses and how the fission products behave and how the 15 containments perform. 16 Therefore, from*that viewpoint, maybe the 17 performance guidelines should be higher than 10 to the minus 18 4. 19 The other problem that he had when he put this 20 together was that it wasn't very clear from the Steering 21 Group's report what the 10 to the minus 4 was. This was 22 discussed today and was discussed at length yesterday, whether 23 it was a core damage type of number or if it was a core melt 24 with no vessel penetration, so he wanted to tighten this up, 25 and his 10 to the minus 5 in both reactor years is a

16 1 large-scale core melt where the vessel in the primary system d 2 has indeed been penetrated and the containment is challenged. 3 You heard numbers like factors between 2 and 10, 4 variations between core damage and core melt where the vessel 5 has been penetrated, depending on who the analyst is. Of 1 6 course, as you all know, it is very hard to precisely identify 7 which are the sequences that will that won't progress from 8 core damage to core melt That depends on the sequence and 9 many other characteristics of the sequence and the system that 10 you are dealing with. But we also know that the PRA numbers the characterization of those numbers 11 are numbers that come 12 are inadequate core cooling. 13 So there are some sequences that if they come from 14 things like station blackout where probably there isn't that 15 much difference between core damage and massive core melt 16 where the primary system has been penetrated. But on the 17 other hand you can have some other sequences where possibly 18 you lose recirculation or something like that, and the chance i 19 of preventing that sequence from going from core damage to l there is a chance to do that with 20 core melt is probably 21 that sequence. 22 MR. KERR: Is the implication of the last line in 23 connection with the rest that although he thinks the 10 to the 24 minus 4 is too large a number, that he would be satisfied with I 25 10 to the minus 5 for an upper limit to a probability of large

{ l 17 1 scale fuel and fission product release? ['~}/ \\, 2 MR. SPIES: Well, he has come up with this number. 3 Let me show you some calculations that have been performed and 4 are very near target. 5 [ Slide.] 6 The 50/50 number that I mentioned earlier was based 7 on 3 times 10 to the minus 4, which really was a 45 to 55 8 percent. If the goal -- if the number is one times 10 to the 9 minus 4, then the probability that in the next 20 years in 10 the population of 100 plants you will have a core melt is 20 11 percent. If you go down to one times 10 to the minus 5, it's 12 really 2 percent. So there is an order of magnitude between 13 this number and this number [ indicating), but if you stop -- 14 if the goal, say, is three times 10 to the minus 5, you are 15 still substantially below the 50 percent. 16 He thinks that the 50/50 number is too high and 17 possibly we should have some other goal that gives you some 18 number that has a better confidence. 19 MR. KERR: I guess I must not have made my question 20 clear. I thought you were saying that the 10 to the minus 4 21 for severe core damage was too big, but that if one adopted 10 22 to the minus 5 for release of the core, melt through of the 23 core through the vessel and release of fission products into ) 24 the containment, that that would be acceptable to him. 25 Is that what the first slide means?

18 ~x 1 MR. SPIES: Basically yes, yes. But, you know, J 2 based on this spectrum of calculations, you know he is not 3 dogmatic that this should be the number, okay. 4 MR. KERR: No, I'm not trying to agree or disagree 5 with him; I just want to make sure I understand him. 6 MR. SPIES: We discussed this last night. Yes. 7 MR. REMICK: A question for clarification. Where 8 does the 3 times 10 to the minus 4 come from? What's the 9 origin of that? 10 MR. SPIES: Okay. This is based on six industry 11 PRA's and in two or three of them the staff did its own 12 review, and this is an average of those numbers from the six bk-13 PRA's. Some of them are industry numbers; some of them have s 14 been revised as a result of the study. 15 MR. LEWIS: Weren't they, in fact, all claimed to be 16 means? 17 MR. SPIES: Yes. At least that's what they claim. 18 But there's always a big difference between what they claim 19 and what the real world is. If you remember, you did the 20 review of the WASH-1400. They had this number that they call 21 a mean and they put a distribution around and then a normal 22 distribution, but that distribution could have been something 23 else. 24 MR. LEWIS: But all those claimed to be means. 25 MR. SPIES: That's all claimed, yes, e

19 -~ 1 MR. WARD: Themis, a question. Do you recall the N_) 2 letter that the ACRS wrote in, I guess, October of last year? 3 I'm not sure it was October. In which we provided an estimate it was an answer to Commission Asselstine's questions. We 4 5 provided an estimate of 6 MR. SPIES: I recall that. I have read so many 7 letters that I don't 8 MR. WARD: Yes, I know how you feel. 9 Well, I just wondered. The number that you seem to 10 be using and that has been floated around as kind of the staff 11 estimate of the probability of an accident in the next 20 12 years is something more like this 45 percent or 50 percent. [ \\m-13 The number we estimated was more, as I recall, like 10 or 15 14 percent. Do you understand what the differences were there? 15 MR. SPIES: The only thing I can that I remember i 16 very well is we came up with this calculation. As I say, it's 17 one minus E to the minus lambda d. If I remember correctly, 18 Commissioner Asselstine took the same numbers and he put some i 19 uncertainties plus or minus, and that number went up to 80 20 percent, and on the other side it went to 10 percent or 5 21 percent. But I don't recall how you arrived at the 10 or 1$ 22 percent. 5 I don't recall 23 Dr. Okrent, you recited yesterday 24 the numbers. 25 MR, WARD: Maybe it's as simple as we used 10 to the

l l 20 f 1 minus 4 as our eyeball average or something -- is an eyeball t\\ 2 average a mean or a median? 3 MR. SPIES: If you used 10 to the minus 4, then -- 4 MR. LEWIS: We used 10 to the minus 4 and stated 5 that we were using the proposed safety goal as the basis for 6 that. Although that didn't specify whether it was median or 7 mean. 8 MR. MARK: But I think we treated it like a mean. 9 MR. LEWIS: Yes. We used it like a mean. 10 MR. SPIES: All I did was use was the paper from my 11 college days of 30 years ago. 12 MR. WARD: They're probably still good. b) \\s, 13 MR. SPIES: So enough of this. This is the issue 14 that he is suffering with more than any other issues right 15 now. 16 MR. OKRENT: Cah I ask a question, Themis? Suppose 17 the Commission were to adopt that proposed alternative. Do 18 you have reason to think that it's approximately achievable, 19 for the large majority of the existing reactors -- and by 20 existing I mean operating or under construction -- by making 21 necessary changes? 22 MR. SPIES: I would give you one view, and I would 23 like most of the experts sitting back there to add or subtract 24 to what I say. 25 I think based on my perception, if we look at the

21 1 sequences that really take you to massive core melt and vessel 2 melt through and do something with them -- you know, like 3 station blackout issue, -- you know, if you can come to grips 4 with such an issue, then I think that the delta between core 5 damage where the stuff is contained inside the vessel and 6 massive melt through the vessel and challenging the 7 containment would be large, or would be maybe a factor of 5 or 8 7, something.like that, And then in that sense, you might get 9 closer to achieving this goal. But that's one view, and I 10 would like to have some of the other experts -- 11 MR. STELLO: I think the correct answer is we just 12 don't know, because you're asking to go back into all of those a p 13 analyses and ask the question, redo the calculations and put 14 them into two bins; those that go to full melt through of the 15 vessel and those that don't, and then, what are the two 16 numbers. 17 I think until at least you have done with the 18 calculations that exist, I don't know that you can answer the 19 question. You need to really know what scenarios there were 20 that went to the full core melt, and what it is that we are or 21 aren't doing anything about. But until you've done that, I 22 don't think you can really answer that question. 23 MR. LEWIS: You will add more uncertainty. 4 i 24 MR. STELLO: You clearly are going to add more 25 uncertainty in the calculation because you are asking the

22 1 technology to be able to discriminate something that I don't 2 think it is yet ready to discriminate. 3 MR. MURLEY: And it is heavily dependent on human 4 action. 5 MR. KERR: The implication, I thought, of 6 Mr. Denton's message was that reactors that complied with his 7 criterion of 10 to the minus 5 for release of large amounts of 8 fission products would be safer than existing reactors. When 9 he says, I am not comfortable with 10 to the minus 4 but I 10 would be comfortable with 10 to the minus 5, either he has a 11 gut feeling or somebody has done some calculations which say 12 that the use of that criterion would produce or would at least Cl 13 convince me that reactors are safer than they are. Now, I 14 must be missing some point of logic. 15 MR. STELLO: You know, we're talking about what it 16 is that Mr. Denton thinks, and I apologize. He promised me 17 he'd be here today. 18 MR. KERR: He has not discussed this with anyone 19 else? 20 MR. STELLO: I have not had any discussion that 21 would allow me to give an answer which I feel comfortable 22 represents his thinking. I'll give you what I view the 23 difference to be, as I see it. ) 24 The 10 to the minus 4 are calculations that we have 25 today that give-us the information we have been measuring \\

23 1 against 10 to the minus 4. S \\_ 2 MR. KERR: Now, the other thing he could be saying 3 -- and I just hate to believe this -- is that if I use a 10 to 4 the minus 5, then the likelihood that I will reach that is 5 smaller, and therefore, things look better without my changing 6 anything. But he's not saying that, surely, is he? 7 MR. STELLO: I doubt it. 8 MR. SPIES: No, I don't think he's saying that. 9 MR. KERR: Well, then, he must be saying that he 10 wants some changes, and that the changes will demonstrably 11 make reactors safer. 12 MR. STELLO: Well, let me give you my opinion. The %,) 13 results that we have today that generally are in the 10 to the 14 minus 4 ballpark, if you look at whether the judgments are or 15 aren't correct that the difference between that number and a 16 full-scale melt is a factor of 2 to 10; if the factor of 2 is 17 correct, then if you adopt 10 to the minus 5 you will be 18 making reactors five times safer. If it turns out it were a 19 factor of 10, there'd be no change. 20 So the answer, as I see it, is if we had infinite 21 knowledge we would know to what extent you would improve 22 safety by the changed language. But I think the intent that 23 he had in mind, in my view, was that there would be an 24 improvement in safety by adopting this rephrased language; 25 that is, reactors would be made safer, that not only would the

24 1 likelihood of a melt-through of the reactor vessel be lower, \\ 2 but the likelihood of severe core damage also would be lower. 3

The not intent being that if you applied this and reactors met O

4 this new standard, that there would be some improvement. 5 Based on what I've heard, I'd say it's between a 6 factor of 0 and 5, in terms of improvement. 7 MR. WARD: One and 5, yes. 8 MR. STELLO: One and five, right, I'm sorry. 9 MR. REMICK: Dave, I have a procedural question. I 10 have a comment I would like to make on this 10 to the minus 4 11 question, but I don't know if you want those now or if you 12 want Themis to go ahead and make his presentation and we can h) (_ 13 come back. So I ask just from the standpoint of efficiency 14 whether you want me to do it now or if you prefer I hold off. 15 MR. WARD: Why don't you go ahead and make it now. '16 I think we're in a general discussion. 17 MR. REMICK: Okay. I think it might be helpful to 18 look at this historically. When the safety goal was developed 19 it was developed from the standpoint that the Commission's 20 responsibility is protecting health and safety. And 21 therefore, the primary goals were public risk guidelines. The 22 idea being that the Commission ought to be able to tell the 23 public that if you live in the vicinity of one of these 24 plants, it's our best judgment that you will not be subjected ) 25 to an increase of deaths from an accident at that facility,

25 1 prompt fatality, greater than one chance in 1000, an ,frq 2 incremental increase. 3 And you should not suffer a chance of a cancer 4 fatality sometime in your lifetime greater than one part of 5 1000, and that was the concept. And then people said, yeah, 6 but you're going to have people come in and say okay, we'll 7 meet that by building the perfect containment vessel. We will 8 worry about ECCS and all that type of thing. So they said 9 okay, well then you should anchor that and make sure people 10 don't put all their eggs in the mitigation basket, but that 11

f. h e y put some in prevention.

There ought to be some 12 quideline, and it was called a secondary guideline, and I (_,/ 13 think it still is, that gives people a feeling for where they 14 should be on the prevention to give them some order of 15 magnitude. 16 Now, I must admit the Steering group is coming out 17 and saying that they think that should be elevated to a 18 full-fledged guideline. So I just wanted to give that J 19 historical perspective on 10 to the minus 4. 20 The other thing is there was a phase in going from 21 NUREG-0880 to Rev. 1 in which this question of what do you 22 mean by large-scale core melt was raised, and it was proposed, 23 and actually it was proposed to the Commission if I recall [ 24 that this be changed to a loss of core cooling capability to (_-} 25 better describe what the intent was, but that was not

26 i accepted. Large scale core melt stayed in. gg 2 But I thought perhaps this historical perspective 3 might help in this discussion. If it is still a secondary 4 guideline I'm not sure it's so important. If it's what the 5 Steering group recommends, that it be a primary guideline, 6 then whether it's 10 to the minus 4 or 10 to the minus 5 or 7 some other number. 8 Now, there's one other thing I'd like to say, I i 9 guess, while I have the microphone. There's a question in 10 mind if it goes to something like 10 to the minus 5, has the 11 staff thought of whether this would require public comment to 12 get full-fledged comment on that point? 13 MR STELLO: Well, what I had hoped was that we 14 would get some sort of discussion as we are having today now 15 on the issue, and try to come out where the staff ought to be 16 in terms of this, and then the question will have to be 17 raised: if you do change, is it important to again provide an 18 opportunity for public comment. But I don't think you want to 19 deal with that issue until we've decided it. I think it's a t 20 good question. 21 MR. REMICK: No, my only point is perhaps it would 22 be good to have your views on that. 23 MR. STELLO: Yes, I agree. I also think that you 4 24 have raised the point -- and I was going to come back to it if you are really interested in changing a plant 25 later l

27 i 7/ 1 performance in terms o 2* safety, the safety is really the \\ 2 consequence of the business. You could have changed the full 3 order of magnitude at the other end, which was quite 4 arbitrarily chosen; there was nothing particularly scientific 5 about a 10th of a percent, and made it a 100th of a percent, 6 and then you would have been able to say you have clearly set 7 as a goal an even tighter standard as that goal for public 8 health and safety because that's the business we're in, 9 protecting the health and safety of the public. 10 That is the other end of this argument. But there 11 is the middle ground that we haven't talked about that I think 12 we need to come to, and that's part of the defense in depth. g 13 It's trying to achieve the balance. And I think that's part 14 of also what was driving Harold, is that middle part of this 15 business that says, you know, -- and Themis made the point 16 earlier -- let's not try to rely too much on the containment 17 performance and the dispersal of the fission products. We 18 have so many other pieces of information that are coming 19 together that make this even more difficult. 20 If the source term results, based on what they look 21 like, -- they also have suggested that the consequences are 22 indeed also lower. So in terms of the actual safety of the 23 public, the research that we have done suggests that plants 24 indeed are safer, which is also the conclusion I think of the 25 American Physical Society than we had previously believed them

l 28 I p i to be. 2 This suggests, again, making them, say, for we are 3 clearly coming to the ultimate question of when have you 4 really achieved " safe enough"? If " safe enough" means another 5 factor of 10, there's a variety of ways to say that, and this 6 is one of them. There are others, however. 7 MR. WARD: Well, I'm glad you brought that up, 8 because I think we have lost sight of the original idea of the 9 safety goal, as you stated it, because what is driving 10 Mr. Speis here to want to lower the number is not a concern 11 about causing more offsite cancers or accidental fatalities, 12 but that he doesn't like the 45 percent chance of a coremelt, \\\\- 13 whether or not it causes offsite fatalities. 14 MR. REMICK: And I'm not sure we should be referring 15 to 45 percent. If it met the safety goal, it would be 20 16 percent. Three times 10 to the -4 is based on a current set 17 of PRAs, an average of those. 18 MR. LEWIS: It would actually be 18 percent. 19 MR. WARD: But it's got nothing to do with offsite 20 effects, except, of course -- 21 MR. SPEIS: This number got notoriety because we 22 addressed the letter to some Congressman, so I put it there 23 for historical continuity and perspective. 24 MR. WARD: Yes, a lot of history today. 25 MR. OXRENT: Well, there are several points that I'd I i i

29 1 like to make. J 2 In the first place, Mr. Denton does make the point 3 that he is not prepared to be as confident about the 4 predictions on containment performance as the PRAs are 5 claiming. So that's one reason for his being uneasy. 6 Let me finish, Themis. I have several points I want 7 to make - -uneasy with this coremelt, and I an equally uneasy, well, so far as history is concerned, just 8 and,.in fact 9 like in the Judeo-Christian tradition, there is more than one 10 time when history began, 11 In fact, before NUREG-0880 began, there was an ACRS 12 document which proposed some trial criteria. There was a mean N 13 1arge-scale coremelt guideline in it of 10 to the -4 per 14 reactor year. It clearly meant melt through the reactor 15 vessel, because there was a lesser hazard stated in it, which 16 showed a rather modest release of fission products involved 17 from the core -- in other words, a TMI type sort of thing 18 and in fact, it suggested that that might be acceptable at a 19 higher frequency. a 20 In the ACRS proposal, there was an effort to achieve 21 " defense in depth" by asking for.both a mean coremelt 22 frequency a'n d a containment performance guideline, given a 23 large-scale coremelt, and that is a part that we are still ' O) t 24 trying to get the Staff to pursue. ( 25 I don't think it's practical to go to 10 to the -6 i i . - - - -,. ~.,., _ _, _ _ _,.., _ _. _ _ _ _ _ _ _

30 1 on coremelt; in fact, my understanding of what transpired at 2 Sizewell B is, the designers there originally thought, we will i 3 try to show coremelt frequency at 10 to the -6 and decided ) l 4 they'd better back off to 10 to the -5, and I think in Italy 5 as well,.to get some credit for containment, to get the 10 to 6 the -6, the serious release outside containment. 7 So I would be quite skeptical, let me put it this 8 way, of anyone trying to show me 10 to the -6, and I don't 9 think I would take on the job of designing the reactor that 10 gave you 10 to the -6 mean coremelt frequency with -- all 11 right, mean -- and it asked them to give it with some 12 confidence, it would be, you knew, obviously even worse. 13 The reason why I asked the original question about 14 10 to the -5 mean is, I guess I, myself, am at this moment 15 skeptical that you will find a factor of 10 universally, 16 reactor to reactor, between the likelihood of an interrupted 17 coremelt accident and the large-scale coremelt. 18 It may occur in some reactors that the dominant 19 event is one where, when you look a t-it with certain modest 20 things or whatever it is, you can decide you can restore it wasn't permitted in that 21 something in the time allowed and you alluded to this 22 particular PRA, but because that because there are a lot of different kinds of 23 yourself 24 scenarios and they vary from reactor to reactor, I wouldn't, I'd be reluctant to count on it, and one of the 25 myself i . - _ _ _ ~... _, _ - _ - - - _ _ _. _ _. _ _ _ _., ~... -

31 /\\~ 1 reasons why I then asked about the 10 to the -5 is, I think it + l 2 would take probably substantial changes I don't know what and let's and even assuring a mean, 10 to the -4 3 kind 4 leave out small reactors like Big Rock -- it may not be ) 5 straightforward, let me put it that way. 6 MR. STELLO: I would like to make a point now, 7 because I think it's important to make right after Dave made 4 8 that comment. I 9 I d o r. ' t believe I've heard Harold say what you 10 said. I read his words again in terms of lack of confidence 11 in the containment. I don't think it says that. and maybe we ought 12 I think that the sentence says 13 to read what he said, because he's not here, rather than 14 trying to put words in his mouth that he may have to write 15 letters to try to get out of them -- let's read what he said. 16 MR. LEWIS: Wasn't he going to be here? 1 17 MR. STELLO: He told me he would be here, yes, and 18 he is not. 19 "The accident prevention guideline is a quantitative i ( 20 corollary of our " defense in depth" concept and should include l 21 a margin to allow for the imperfection in the current methods t 22 of predicting coremelts, fission product behavior, and i' 23 containment performance." 24 That's what he says, and I don't believe that's what i 25 you said he said, so I think we ought to rely on the written I

3 l i 32 0<. 1 words. 2 MR. OKRENT: I am perfectly willing to stand with 3 what you read, because it may be 4 MR. STELLO: Okay. One more point, since I already 5 ' rudely interrupted, and I apologise. 6 If you talk 10 to the -5, Dave, you ought to 7 understand that there's roughly a half a dozen to a dozen 8 scenarios which make that up, which means ycJ are talking l 9 about 10 to the minus -6 scenarios that ycu are trying to deal 10 with, and you are also getting pretty far out in the spectrum 11 in light of the uncertainties, and you may even spill into the 12 10 to the -7 scenarios. I 13 And you are asking this technology to do, I think, 14 maybe a touch or two more than really it's prepared to do. j 15 MR. WARD: Well, I am concerned about the same i 16 question. 17 MR. WARD: Carson? 18 MR. MARK: I wanted to add one thing to Forrest's Y j. 19 history. The secondary criteria, they didn't make enough of 20 it in the Steering Group report, but the real fact is that 21 that's the only criterion on which they can lay their hands 22 and make any conceivable application. It isn't a secondary i 23 criterion; it's the only operable one. 24 MR. LEWIS: I just want to comment on Vic's 25 ,ereading o, aaroid.s basis and remind us that, i, i remember

33 1 the words correctly, he said to allow for imperfections in the ,Y 2 -calculation. That is a basically conservative position. It 3 doesn't say allow for bias in the calculations, but to allow 4 for imperfections. So he wants to move a criterion down which 5 should be realistically calculated to provide conservatism 6 against imperfections in other calculations. And that's a 7 generic disease, that these things have to be done 8 realistically. 9 And I read Harold's letter as betraying the routine, 10 every where around here, conservative bias applied to 11 realistic criteria. 1 12 MR. WARD: Themis, did you have anything else now? f this is really the main thing, 13 MR. SPEIS: I have 14 you know. 15 MR. WARD: Okay. Well, just go ahead. I think we 16 sort of interrupted you. 17 MR. SPEIS: I have a second Vu-graph now. 18 CSlide.] and 19 Some of the other comments that you have read 20 Frank explained them very well yesterday, so ! can't add more 21 to what he said -- we do agree with the Steering Committee's 22 recommendation that averted onsite losses be included. As 23 Frank said yesterday, since this issue has been discussed very () 24 extensively and there are all kinds of ideas and nuances and 25 statements, we suggested that, you know, maybe a separate

l 34 i document should be put together and articulate in great detail -s U 2 all the pros and cons, whether it's economic regulation or 3 safety regulation or under or over-regulation. That's 4 basically what he said. 5 The other thing we are uneasy with to some extent 6 are the implementation guidelines. I have a separate Vu-graph 7 8 MR. KERR: May I ask a question, since I think we 9 are going to a different topic? 10 You say you agreed to the safety goal Steering 11 Group's inclusion of onsite. Suppose as an alternative, one 12 said, "All reactors must meet the 10 to the -4 guideline," (,f 13 period, and you would use ALARA to improve on that. 14 Would you feel the same way about including onsite 15 costs? What I'm asking, is the onsite cost simply a lever to 16 get to the 10 to the -4, which is where you think things 17 should be, or is it a matter of principle that you think that 18 the onsite costs ought to be considered. 19 MR. SPEIS: Well, I think as a matter of principle, 20 all costs should be displayed. 21 MR. KERR: Now wait. Let's not talk about "all 22 costs," because we'll never be able to -- we can talk about 23 onsite costs, maybe. 24 MR. SPEIS: Well, all important costs, all 25 important, as someone said yesterday. And when we talk about

l 35 7"%g 1 onsite cost, what are we talking about? Averted radiological t i v 2 cleanup and replacement of power? 3 MR. KERR: Okay. So it's a matter of principle, and 4 not just an effort to get plants to the 10 to the -4, say? 5-I'm not trying to be critical. I'm just trying to understand 6 how the Staff reached the conclusion, because, you know, it 7 helps me to consider it, if I understand it. 8 MR. SPEIS: Well, everybody always, you know -- 9 their motives are not 100 percent pure. Maybe there is a 1 10 percent impurity in Harold's motive. If I get this, then I 11 can 12 MR. KERR: I don't see anything impure about 13 wanting to have a lever to accomplish something you think 14 should be accomplished. 15 MR. SPEIS: I think on this one, I will ask the 16 specifie question of him to make sure that I don't speak for 17 him. But I think it's a matter of principle, but as I say, 18 uaybe that 1 percent impurity is always important. I don't 19 know. 20 MR. KERR: Thank you. 21 [ Slide.) 22 MR. SIESS: If it were a matter of principle, would 23 you have any problem if it showed you that 10 to the -3 was 24 okay? s 25 MR. SPEIS: Excuse met

36 1 MR. SIESS: If it were simply a matter of principle, -s v 2 would it give you a problem if the cost / benefit analysis 3 showed you that 10 to the -3 was okay? 4 MR. OKRENT: Exouse me, Chet. I have to ask a 5 question of you first, because if they say there's a criterion 6 of 10 to the -5 that should be met, then I don't see how 7 cost / benefit can show that something less than that is okay. 8 MR, SIESS: I agree. But I don't know why you need 9 to make cost / benefit at all, if you've got 10 to the -5 as a 10 limit. 11 MR. OKRENT: But you see Harold is proposing a speed 12 timit or whatever you want to call it of 10 to the -5. So if 13 that is adopted, then what s 14 MR. SIESS: Well, by " principle," maybe he meant 15 that if you are going to use cost / benefit, then you should 16 include all costs and all benefits. But I was assuming that 17 there was a matter cf principle involved in using cost / benefit 18 it. Maybe I mistook it 19 MR. KERR: Well, your comment puzzled me a bit, 20 because I thought, given that, let's suppose we had a 10 to 21 the -4 goal. I thought that oost/ benefit was to be used to 22 determine whether plants that did not meet the 10 to the -4 23 must make changes to meet it, and you made that decision based ) 24 on a cost / benefit analysis; is that not correct? i 25 MR. OKRENT: Well, there are different ways in which

.____.m.. 1 i 37 4 let's talk 1 one can use cost / benefit analysis, and one is I 2 about 10 to the -4, and let's say a plant doesn't meet it, and i i 3 you can say, "I can't find any cost-effective way to get from 4 10 to the -3 to 10 to the -4; therefore, I won't make any 4 5 changes." That could be one way to use cost / benefit analysis, i 6 Another is to say, "10 to the -4 is our design i 7 objective, and we intend to meet it, if it's practical." Then i 8 you find the most cost-effective way, if you will, of getting 9 up to 10 to the -4, but 10 MR. XERE: But regulatory people don't have to find 1 11 the most oost-effective way. That's left up to the designer. I 12 MR. OKRENT: But it may still have a ratio which is I 13 -- where the cost is larger than the benefit. But you said, 14 "This is a goal that we intend to meet, to the extent one is l 15 able to calculate it and'so forth." 1 16 MR. ROW 50ME: I wanted to insert a couple of points. t I 17 MR. KERR: ! yield to the Senator from -- 14 CLaughter.3 19 MR. ROWSOME: Cost / benefit already has a l 20 well-established role in this agency in at least three arenas 4 in the backfit policy, in the regulatory analysis of new l 21 i I 22 generic reactor safety standards, and in NEPA. The safety 4 ) 23 goal arena has been chosen as the arena in which the agency 24 will codify how it implements that. But it has a policy \\ i l 25 existence quite apart from implementing the quantitative 1 l i i ,n _

38 1 design objectives in the safety goals. O 2 No one in the Staff has entertained the idea of 3 making any coronelt frequency guideline a requirement that all 4 plants must meet, so that some of these comments the s 5 discussion has gotten rather academic and far afield. ] 6 MR. KERR: As Mr. Lewis pointed out this morning, ! 4 7 don't consider " academic" a pejorative term. 8 [ Laughter.3 9 MR. ROWSOME: I don't either. I am just pointing 10 out that we talking about applications of cost / benefit in 11 arenas other than just implementation of the quantitative 12 design objectives. ^ \\ ,) 13 MR. WARD: Well, I heard that. But for some reason, 14 we seem to have the same fundamental difference of opinion 15 that we had yesterday between -- ! think it's between Bill 16 and Dave, and I am amased. 17 MR. KERR: I never disagree with Dave. 18 MR. OKRENT: And I wasn't offering an opinion. I 19 was trying to say, I have seen cost / benefit used in different 20 ways. 21 MR. WARD: Yes, but maybe we are about to hear how 22 NRR proposes to use it. Is that what we tre about to hear, 4 23 Themis? ( 24 MR. SPE!S: Well, I was going to bring the point 25 that we made, illustrated graphically. It will take a minute

39 rg i or less than a minute. U 2 MR. WARD: Okay. 3 MR. SPEIS: You are all familiar with the 4 implementation to exclude consideration of regulatory actions 5 between 10 to the -3 and 3 times 10 to the -5, unless -- but t 6 you check the mortality goals, of course, and if met, then you 7 check whether a specific sequence exceeds 10 to the -5. 8 I was going to make the point that, you know, when [ 9 PRAs are done and we have to review them, there is always the 10 problem that, you know, the analysts, they can play games with 11 the sequences. They can combine them this way or that way. 12 And I have an example at the bottom of the page here that \\ms/ 13 maybe somebody will come up with 100 sequences, and all of 14 them will be slightly below the 10 to the -5f, and when you 15 put them together, you come up with 9 times 10 to the -4, 16 which will be slightly bel'ow this 10 to the -3. 17 So this goes to the heart of uncertainties in how 18 one performs this calculation and what the reviewers accept l 19 and don't accept. So our point is that those things are not 20 very well articulated in this implementation. But maybe we 21 have to do more by experience or by some other vehicle. 22 MR. SHEWMON: Back to Bill's comment, or the use of 23 cost-benefit, I know I read some place, though whose position l 24 it was, I don't know, that there was a window some place 25 around 10 to the minus 4 to 10 to the minus 3 in which l i

40 i cost-benefit would be used, and above that you fixed it no V 2 matter what, more probable, less probable, something else. 3 !s that still 4 MR. SPEIS: If a specific sequence has an arrival I 5 rate of more than 10 to the minus 5, then you look it over and 6 make sure that there isn't something to be done. 7 MR. KERR: Is the window that Prof. Shewmon talks e about between 3 times 10 to the minus 5 and 10 to the minus 3, f 9 that's the window I believe to which you refert 10 MR. SHEWMON: And it was only that window that 11 cost-benefit would be used. Is that still the policy or 12 proposal? Whose proposal was it? 13 MR. STELLO: Let me say what it is that we are 14 proposing as a rule, and the rule before the Commission right 15 now, basically says if ever you conclude that a plant is not 16 adequately safe, that is that you can reach the conotusion 17 that there is no risk to the public health and safety -- if 18 you reach that conotusion you don't do cost-benefit. You 19 impose it. 20 If it is a matter of compliance, the proposed rule 21 that is before the Commission says you don't do it. 22 What Themi has in front of you is the part of the 23 implementation program for safety goals and how to try to use 24 that as a way in which to develop a more meaningful process to 25 try to deal with some of these questions.

41 1 Perhaps in 10 years we will be able to then change 2 the language I just gave you into a more quantitative form. 3 The language I gave you is the language of the regulations. 4 That's the requirement. 5 Let me remind you again, let's not continue to 6 characterise these safety goals as requirements or as 7 standards. They are not meant to be that. They are not going 8 to be used for that purpose. 9 MR. SHEWMON: Well, if I used safety goal, ! 10 misspoke. 11 MR. STELLO: Well, I gave you what the rules are, 12 and the rule that is now proposed and before the Commission. O(_f 13 As ! recall, that is in fact 14 MR. KERR: But it seemed to me there seemed to be 15 exceptions to what you said, and perhaps you can explain why 16 there are no exceptions. There was a branch technical 17 position, for example, that had to do with behavior of 18 auxiliary feedwater systems. I have been told in the past 19 that branch technical positions are not rules. One could 20 satisfy all the rules without conforming, therefore, with this 21 branch technical position and yet I think new reactors at 22 least are being required to conform to that branch technical 23 position. 24 So when you tell me that as long as they meet all s 25 the r t:1 e s, nothing more has to be done, I am pussled.

42 1 MR. STELLO: I have just given you what is in fact 2 the requirement. The way in which we impose requirements 3 takes on a new meaning when you talk Staff positions. What 4 are all the Staff positions? 5 Now the Staff positions are those things that are 6 embodied in the standard review plan, and we tell a plant now 7 if you wish to get a license, show us how you conform to the 8 standard review plan. 9 Now listen to the next sentence. It's very 10 important. 11 If you choose not to follow the standard review 12 plan, you wish to deviate, tell us where you wish to deviate \\ 13 and then tell us how you meet the basic underlying regulation. 14 MR. KERR: Vio, we're playing games. 15 MR. STELLO: No, we're not. It's very, very 16 important. What the requirements of the Commission -- they're 17 what they are. They're our only legally binding requirements 18 and obligations of the Licensee. 19 The way we do our review process is a standard 20 review plan, whleh you've seen, is a very thick document, and 21 we basically tell the Licensee do it this way, and if you 22 don't, you have got to be able to justify why not. 23 MR, XERR: And the Itkelihood of his being able to 24 do that is what? 10 to the minus 6, 10 to the minus 7, and he 25 knows it?

~_ F 43 1 1 MR. STELLO: No, I think the Beaver Valley contest g-1 \\ 2 has shown that of those areas where he raised the contest 3 think, if my memory serves me right now, about 12 of the 16, j i 4 the Licensee, we agreed with him and he departed from the i 5 standard review plan. 6 Am I right, about 12 out of 16? 7 So the answer to your question, the most recent i 6 example I have is where the contest was raised, 12 of the it, l 9 and they all aren't decided yet, the Licensee departed from 1 ) 10 the standard review plan and the Staff accepted it. 11 So the real world, if we can make it ever work i l 12 sensibly and manage it correotty, it ought to allow for that. s_/ 13 To allow to deviate from these practices when there is a case i 14 made to do that, j r l 15 MR. LEWIS: In that case, did the Staff accept it i f j 16 cheerfully? t i 17 MR. STELLO: I never asked. I don't know, i to MR. LEWIS: It took a long time. 19 MR. STELLO: It did indeed, because, you know, we i 20 began a process, it was the beginning of a new way in whtoh to j 3 21 manage backfit, and it was just beginning. That process has f 22 only been in place about a year. i 23 Themi, I would suggest that in light of the opening l 24 comment that we are not going to discuss mean and median, and [ i 2S ! plead with you do not, so the next slide I would not use. I = w.e- --c., .,--...----e...,_.--..y----..y,- y-w,--- m -w, y, w,,.,,3-

44 1 MR. SPEIS: I have brought an additional expert with s 2 me to help in the discussion, Mr. Thadani. 3 MR. STELLO: We are not going to discuss it. 4 MR. XERR: I thought we had a caucus just before 5 lunch and decided that we weren't going to talk about it any 6 more today at least, didn't wet 7 MR. WARD: Yes. 8 Let's see, Dave, we've got, you know, 45 minutes 9 more. 10 MR. OKRENT: Is there a representative of 11 Mr. Minogue here? 12 MR, STELLO: We made a call to Mr. Minogue's 13 office. I don't know whether his representative is here yet. 14 1 guess not. He is on his way. 15 But let me, before you move to that, if I could ask 16 Themi one question. 17 MR. SPE!S: I want to make one more point. Harold 18 told me to stress to the committee here that he will continue 19 to write his views to Diroks and to the Commission and to the 20 other offices and to the ACRS, even before an integral view 21 has been packaged by the EDO, because, you know, even though 22 he feels that maybe his letter was misquoted and he received 23 !arger notoriety, but he still told me to relate to you that's [) 24 the way he will always do business. v 25 He wants to express his views because maybe in some

45 4 fw 1 circles the letter was mischaracterised. I am not referring ( i 2 to the ACRS. 3 MR. STELLO: In order to help, Themi, let me ask you j 4 a question. Based on my understanding of discussions with i 5 Harold, it is my understanding with Harold that one issue he 6 thought needed more discussion and ventilation is the 10 to 1 7 the minus 5 issue, and given that that got the proper i j 8 ventilation, that he was prepared to support the steering 9 committee report; is that correott 10 MR. SPEIS: Yes. 11 MR. STELLO: Okay. Thank you. 12 MR. OKRENT: Well, anyway, I must say I hope the 13 Commissioners and the EDO encourage their senior staff to 14 express their strongly held or important opinions, and that 15 there isn't anything done that would diminish this expression 16 of opinion. 17 MR. STELLO: 1 agree with you, with the exception of i 18 this parttoular sequence that we have had with the ACES, where 19 we tried very much not only to get Harold to put down his 20 comments, but to bring him down and discuss it with you. 21 1 don't know that this experience has been one which 22 would warm the cookies of one's heart to do it repeatedly, I guess 23 MR, KERR: Vic, I must 24 MR, WARD: Well, did he have a bad experience at ( 25 ACRS? What are you talking about?

46 f-w i MR. KERR: Yeah, what's the problem? 2 MR. STELLO: I sat in front of -- well, I guess 3 behind you-all and I listened to the characterisation of the 4 Staff, in the way in which the safety goal issue has been 5 handled, and some of the comments made about what the Staff 6 has done and where it is, and I don't think they were 7 complimentary. 8 MR. KERR: Oh, I thought you were talking about 9 Harold's letter. 10 MR. STELLO: Harold's letter being one part of it. 11 That's part of the controversy, 12 MR. KERR: I did not hear a criticism of Harold's (s 13 letter. Lewis said something about being complimentary, 14 MR. LEWIS: Well, this is a serious issue. 15 MR. STELLO: ! think it is. I think we are coming 16 down here to try to get discussion and advice, and we weren't 17 finished by a long shot, and I think some of the discussion up 18 there, you know, left me haunted. But I too will not hesitate 19 or shy away from coming back to the committee when we are not 20 decided. 21 MR. LEWIS: No, I'm not talking about you, but ! 22 hope you are not telling us that there are elements of the 23 Staff that will not speak to ACRS unless they get favorable ( 24 reviews from ACMS. 25 MR. STELLO: Oh, I don't think anybody on the Staff

47 N 1 has that view. All of us have been around too long. 'V 2 MR. LEWIS: Well, wait a minute, now. You mean you 3 didn't mean what you just said? 4 MR. STELLO: I mean what ! Just said. 5 MR. WARD: Okay. Well, let's go on to the next 6

speaker, 7

MR. OKRENT: Okay. I also hope that the EDO will 8 not, on important issues, take steps that would diminish the 9 kind of technical interchange that has been possible, in fact, 10 in large part because of the presence of serious thoughts by 11 various members of the senior staff, 12 MR. STELLO: Dave, let me make it very clear. The 13 EDO sought out those views. He asked for them in writing, and 14 he wanted them, because he thinks it's very important to get 15 those views from his senior managers, and he thinks it's 16 important to have adequate discussion and ventilation of them 17 before he decides issues. 18 MR. OKRENT: That's fine. 19 MR. STELLO: So let's make it absolutely clear on 20 the record that not only does he support and does it, and he 21 will continue to, I assure you. 22 MR. OKRENT: Well, we are in perfect agreement. 23 1 have a small question for Mr. Rowsome. 24 MR. KERN: While we are on this issue, Vic, 1 don't 25 know what significance this statement will have, but I want to

r a [ 48 1 make it. 4. 2 Just as the AbkB is not of one mind on many issues, 3 1 don't think you should take comments by individual members 4 of the ACRS as representing ACRS views. 5 MR. STELLO: I know better than to do that also. 6 MR. KERR: Okay. 7 MR. OKRENT: Frank, when you were standing, you said 8 something I didn't quite understand about 10 to the minus 5, "9 and let me pose-an iffy question to you: ~ 10 If in fact the Commissjon did adopt the suggested l 'i - alternative in Harold Denton's memo,'how would you envisage pursuit, since it's 12 its -- well, I won't say implementation ( 13 an objective? 14 MR. ROWSOME: Let me start out by. answering that by 15 ' stepping back to the discussion we had about half an hour ago 16 about how accurately you can measure that"," or whether you can 17 measure it. 18 The credit g t, v e n in PRA for corrective action in 19 repairing faulted systems is now calibrated to what we believe 20 to be the time of core,uncovery. If we calibrated it instead 21' to the time we believe a core meltthrough would occur, that 22 estimate would be no more nor less accurate than it is now. 23 In fact, there are a lot of sequences in which ( theorettoally possible sequences in which equipment is 24 23 nonfunctional by virtue of cognitive era} ors by operators, or

i 49 1 the system is operable, but the operators are unaware that it O 2 is not functioning, perhaps because of faulted control s 3 instrumentation and the like, so that in fact -- and those 4 sequences are normally not considered or not effectively 5 predicted in PRA. 6 So I think the case could be made that one could be 7 more accurate in predicting the frequency of vessel 8 meltthrough than one has been in the past in PRA, in 9 predicting the frequency of severe core damage. 10 Now as to implementation, there are many ways one 11 could contemplate implementing it. Let me set you an 12 example. The FAA sets a target accident frequency for a 5 13 particular make and model of aircraft on the basis that there 1 14 is a fairly small percentage chance, less than a 45 percent 15 chance, that any one of those planes will have a serious 16 accident in its whole service life. { 17 They know they cannot in fact realise that. The i 18 frequency of human error giving rise to aircraft accidents is 19 about two orders of magnitude more frequent than that goal. 20 Nonetheless, it is a goal, and where they do know how to 21 implement it, they do. 22 So we could, without question, on some sequences, on i 23 some classes of sequences, some classes of vulnerabilities, 24 make intelligent use of that goal. Where there are particular ) 25 !arge uncertainties, as is frequently the case with external l . -, - - -, -. -., ~,,,, - -, - ---..n,----- ,-n,---.,,n

50 gg i events, one simply doesn't know whether one has met that goal 2 or not. s 3 I do not anticipate and would not recommend that the i 4 Staff go ultraconservative and start moving heaven and earth 5 trying to beat down those frequencies. I don't see the Staff 6 doing it, and I don't think it's necessary or appropriate. 7 On the other hand, I think it does make good sense 8 to target a frequency that we can really be comfortable with, 9 and that does set as a goal some improvement on where the 10 historical experience would suggest we are today. 11 MR. WARD: Let's see, Frank. You said there was not you didn't think there was a difference in the uncertainty 12 13 in the ability to recover before the point of core meltthrough 14 and the uncertainty and ability to recover before the point of 15 core uncovery? Is that what you said? 16 MR. ROWSOME: Well, obviously there is a little 17 phenomenological uncertainty having to do with how long a core 18 has to be uncovered before it's likely to go through the 19 bottom of the vessel That will likely be uncertain to within 20 10 or 20 minutes on the basis of the way we can do 21 phenomenological analysis now. 22 MR. WARD: But I thought there were important 23 questions about adding water to an overheated core. In the () 24 case where the core is just beginning to uncover, I guess it 25 is fairly clear that simply adding water at the maximum rate

51 1 is obviously the best thing to do. With a core that is s 2 severely overheated and beginning to melt, I thought there 3 were some concerns that that may not be the optimum thing to 4 do. 5 MR. ROWSOME: Whether those would ever appear as 6 important in the overall frequency profile of the plant 7 depends on the nature of the sequences. There are some 8 sequences in which you know once you have entered them, you 9 are intrinsically beyond the point of no return. Vessel 10 rupture, gross interfacing system LOCA, sequences for which 11 there is no potentially operable system that could turn around 12 core damage. And so those problems don't enter at all. tN 13 There are other sequences in which you know you have 14 some core cooling throughout. The sequence by definition 15 entails 50 gpm or 100 gpm or something like that into the 16 reactor vessel, and there may be some marginal questions on 17 some of those depending on exactly what the flow rate is, 18 whether you might go through the vessel or not, but those 19 would be a distinct minority sequence; 20 That kind of sequence has not proved to be important 1 21 in PRAs in the past, Usually risk is governed by sequences in j 1 22 which all core cooling has failed outright. But there are i 23 some borderline sequences where whether in fact you go through ( 24 the vessel at all regardless of operator action is in doubt, 25 sure.

52 1 But I think on balance for most sequences, you have g-%g 2 a disabled but theoretically capable system, and you are 3 trying to estimate the likelihood that operators will manage 4 to jury-rig them into operation, even though they did not 5 start on demand for some reason or failed in service. 6 And the uncertainties associated with that are 7 typically orders of magnitude. And I think that overwhelms 8 the phenomenological uncertainties involved in this in 9 practical applications. 10 MR. KERR: Frank, you said something in your earlier 11 comments which I interpreted to mean that you thought all 12 existing reactors or the population of existing reactors 13 needed to achieve a significant improvement in safety, but I 14 wasn't sure that was what you meant. 15 MR. ROWSOME: I said I thought it would be sensible 16 to set goals that implied improvement. I do not, as I have 17 said to this committee before, believe that commercial nuclear 18 power plants on balance pose undue risk to public health and 19 safety, and I continue to feel that they do not. 20 MR. STELLO: I guess I understood his comment in 21 light of the airline analogy is that we clearly ought to set 22 the goals beyond where we think we are going to achieve them, 23 just to have them as goals, as aiming points, rather than -- O) 24 MR. KERR: I was asking for clarification because I gv 25 wasn't sure I understood him. If that is what he is saying, I

53 1 understand 2 MR. STELLO: That is what I was going to ask. Is 3 that what you intended? 4 MR. ROWSOME: Yes. 5 MR. WARD: Well, we heard these described yesterday 6 as aspirational goals, and I thought the response yesterday 7 was that clearly the goal we are talking about here or have 4 8 been talking about for two years is not an aspirational goal, 1 9 it is a goal that we expect the plant to realize. So I think 10 I just heard something different from that. 11 MR. KERR: I think I just heard something different, 12 too, but at least it was clearly stated and understandable. f% ks 13 MR. WARD: That's right. Okay. m 14 MR. STELLO: I don't know that I ever remember the 15 goals ever being characterized more than targets or aiming 16 points rather than in any sense a standard. I will say, 17 though, that I think we ought to recognize that once we come 18 out with this safety goal, should we ever come out with one, 19 that there will be a significant force that will develop that 20 will want or move to have all reactors meet whatever these 21 goals are, exactly, at least. 22 MR. LEWIS: Didn't the Commission two years ago when 23 it first started make an explicit statement that we do not () 24 expect every plant to achieve these goals? 25 MR. STELLO: That's correct.

54 1 MR. LEWIS: I am not disagreeing with your 2 prediction. 3 MR. WARD: Okay. Anything else? Dave, do you have 4 anything else? 5 MR. OKRENT: Well, we have a choice. I guess at 6 the moment there are no further comments from the Staff; is 7 that correct? 8 MR. STELLO: Well, I was hoping that we could 9 continue to find a way to talk longer and that our researcher 10 would be here. He indicated he was on his way. 11 MR. WARD: Well, it looks like they are not going to 12 be. We are already behind our agenda time. 13 MR. OKRENT: We are not behind the allowable agenda. 14 MR. STELLO: He is here now. 15 MR. OKRENT: Well, why don't we then hear from 16 Research and Denton. Let's see. When did you reconvene the 17 meeting, a quarter of? 18 MR. WARD: Yes. You have got another half-hour. It i 19 owuld be nice to make up.a little bit. 20 MR. OKRENT: We can make up some time. I would 21 rather make up some time now and use it when we are going to 22 review the letter tomorrow or whenever it is, next up. 23 Frank Gillespie is in Research. He is the Head of ) 24 the Division of Risk Analysis and Reactor Operations, or 25 something like that.

55 es 1 MR. WARD: And Human Factors. 2 MR. GILLESPIE: I apologize. We thought we stated 3 our position yesterday, and inadvertently, due to another task 4 force I'm on, I didn't come down today. 5 The Research position is fundamentally in support of 6 the group's report. The one comment that we wrote in in May 7 on the group's report has been, at least in substance in the 8 reading of the new paper, our concern on defense in depth has 9 partially or maybe totally -- because I only got the paper 10 yesterday -- been addressed in the new paper that has been 11 developed by Vic's group. 12 MR. OKRENT: Which? We don't have the benefit of it, w \\ s,/ 13 or at least I. don't 14 MR. GILLESPIE: Oh, I got it from here. 15 MR. STELLO: Excuse me. Yes, you do. 16 MR. GILLESPIE: The draft Commission paper. 17 MR. STELLO: The one that we referred to as the 18 draft is the one that he is talking about, which I assure you 19 he has. 20 MR. GILLESPIE: Yes, you had it much longer than I 21 did. It's dated June 19th. 22 MR. OKRENT: Oh, I apologize. 23 MR. STELLO: We accept. [ 24 MR. GILLESPIE: I only saw it yesterday, which 25 reacted to some of our concerns. Our concern was -- I think in

l 56 (g 1 our letter we said this. We are not concerned with 2 cost-benefit. We are more concerned with the idea that 3 defense in depth not be lost, and we fully support that. I 4 think in the committee's report in the examples that we used 5 in there, there was some emphasis put on the case studies on 6 the fact that you really had to consider the uncertainties in 7 some of those case studies to have them fall into a positive 8 answer under the safety goal proposal. 9 I think you really have to consider that, and that 10 is where we were coming from yesterday. So our concern was 11 one that you don't want to lose the ability to make things 12 better just because a plant meets 10 to the minus 4. There [ t 13 may be safety fixes which really are safety effective below 14 that. 15 MR. LEWIS: Could I just understand that comment? 16 You surely don't mean that you don't want to not consider 17 improvements in safety when the plant is quite safe. You said 18 almost that, and I think you didn't mean it. This is the open i 19 endedness of safety improvement that I'm concerned with. 20 MR. GILLESPIE: Yes, and this is what I said 21 yesterday. I don't have in mind a structure on how to put an 22 entirely closed end on it, but indeed the safety goal or the 23 way the paper, as I understand it, reads now from quickly ( ) 24 looking at it, you know, it identifies this is something that 25 would be used in parallel with the normal staff the way we

57 's 1 have done it in the past. It would be used as another 2 yardstick. 3 MR. WARD: Do you just mean ALARA? Is that all you 4 are saying? 5 MR. GILLESPIE: No. I don't know if I am saying 6 ALARA in the terms that it is written up. I guess I am saying and someone asked 7 ALARA in the sense that if there is a fix 8 yesterday about do you have a list of fixes. No, I don't have 9 a list of fixes in mind. If there is a fix whose significant 10 safety benefit could be shown, even though the plant met a 11 particular core melt frequency safety goal, the policy should 12 .have enough flexibility to allow the Staff to show that b\\ss 13 significant increment in safety is, in fact, worth it; and I 14 think the paper now says that. 15 MR. SIESS: Is it always worth it, Frank? Is there 16 any point below which you wouldn't go even if it were 17 cost-beneficial? 18 MR. GILLESPIE: Well, if it is that diminimus, I 19 don't think you could make the case that it was a significant 20 safety fix. 21 MR. SIESS: I know, but there are lots of 22 definitions of diminimus. Some people wouldn't think 10 to the 23 minus 5 was diminimus. ) 24 MR. GILLESPIE: If there was -- 25 MR. SIESS: So I am asking: you must have some

58 1 diminimus in mind, at least by your judgment. 2 MR. GILLESPIE: The diminimus would depend on the 3 benefits you think you are going to gain from it. 4 MR. SIESS: Well,.irt that case you would keep going 5 as long as the incremental benefit exceeded the incremental 6 cost; right? 7 MR. GILLESPIE: As long as the incremental benefit 8 exceeded the cost. But I mean cost in terms of other than 9 necessarily just dollars and sense. i 10 MR. SIESS: Okay, as a total incremental benefit and 11 total incremental cost, you would keep going, and there would 12 be no diminimus. There would be no how safe is safe enough O 13 and safe enough when it doesn't cost anymore as long as you 14 started low enough. I 15 MR. LEWIS: This is an important issue that is being 16 raised. 17 MR. WARD: I don't understand this. This is either 18 ALARA or unrestricted ratcheting. I haven't figured out what 19 you are saying. Will you pick one of those? 20 [ Laughter) 21 MR. GILLESPIE: What we have not done is attempted 22 to put a diminimus quantity on. If you want to call it -- 23 there is a point when you are not going to be able to make a s 24 case that a safety benefit is worth whatever the cost of doing 25 it is. Now, if I don't have a specific safety benefit in mind

59 1 and therefore I don't have a specific cost of installation in g '( 2 mind, I can't tell you what diminimus quantity that particular 3 fix would have to cross to do it. 4 MR. KERR: In your consideration of this issue, do 5 you give some thought to the fact that when you get down below 6 10 to the minus 4 or 10 to the minus 5 or somewhere, you are 7 in a region of uncertainty in which not only is the core melt 8 frequency uncertain, but the effect that you have on it by 9 doing something to the plant is also uncertain, and indeed, 10 the SION maybe uncertain? 11 MR. GILLESPIE: In which case showing a definite 12 safety benefit is going to be very difficult, so I am agreeing 13 14 MR. KERR: Well, I agree showing they have any 15 benefit at all -- 16 MR. GILLESPIE: "Is difficult. 17 MR. KERR: And indeed, one may have a considerable 18 amount of concern that doing something makes things worse. It 19 has happened, you know. 20 MR. GILLESPIE: Yes, it works both ways. 21 MR. EBERSOLE: I am going to drag out an ancient but 22 classic case to illustrate the point. For many years there 23 has been a flap about the 10-inch high pressure steam line I\\ 24 that feeds the HPSI pumps at the Brown's Ferry and similar 25 plants. If the pipe bursts on the outboard side of this, the aw -- ,e

60 rx 1 potential is to inject steam from a 10-inch main into the U 2 entire three-unit station and produce appalling damage 3 effects. 4 The thesis that prevents that is that the valves, 5 the isolation valves will close if such an event takes place, 6 and even that event itself is of low probability. But little 7 is known about the reliability of those valves to do that. 8 The operation under dynamic loads is poor. The original test 9 was poor, the post-operational testing and duration testing.is 10 virtually negligible, and we just simply don't know whether 11 those valves will intercept that flow. 12 At one point in the history of the plant, one of b 13 these very pipes was attempted to be straightened by a water 14 plug that raced through the system. The valves did shut. Had 15 it not, it was an inviting place to start a three-unit 16 catastrophe. 17 The cost of fixing this is not very much. You just 18 close the valves in the first place and leave them closed, and 19 then you open them when you need them. The cost is silch. I 20 Yet it is not being done because of the rigidity of the 21 ideology we have in the regulatory process. 22 MR. STELLO: May I try? I don't know that I am 23 going to help, but let me use this example to illustrate that 24 in the papers you have, we have tried to say there are safety 25 goals that we are trying to develop, but there is a regulatory

61 e^ 1 process. I think Frank is trying to talk about both sides of v 2 these issues. 3 If the traditional regulatory process -- and I will 4 use Jesse's example. Say that the wisdom is that those 5 valves ought to be closed and you don't have a detailed 6 comprehensive cost-benefit analysis that shows that they, in i 7 fact, ought to be closed, although I think perhaps one would 8 show that they ought to be as well, but let me just. assume 9 that they didn't, and Frank is, I think, suggesting that we 10 aught to be able to use the traditional regulatory wisdom to 11 make a decision where we believe that that particular decision 12 will substantially enhance safety, even not without the (~h \\ 13 benefit of a comprehensive or a cost-benefit analysis at all, s-14 using traditional methods. 15 The paper does, in fact, say that that is what we 16 are going to do. We are not supplanting the traditional 17 methods with safety goals. Where there is such a case and it 18 does arise in conventional wisdom of the Staff and coming to i 19 the ACRS and getting the regulatory advice, the best we know 20 how, that says an action is an appropriate action to take, we 21 ought to take it, even though the cost-benefit analysis may 22 not even be available. 23 I think that is the point that Frank is making. ( 24 MR. GILLESPIE: Yes. It's a parallel process. 25 MR. SIESS: Well, that's clear enough until you .,-..r-

62 i start confusing us with the cost-benefit analysis. 2 MR. STELLO: Did I help to straighten it out, Frank? r 3 MR. GILLESPIE: Yes. That was it. I think if you t 4 read our comment letter, Vic articulately put one particular 5 sentence into the paper from it which said we thought the j 4 6 cost-benefit issue was just kind of I forget the exact 2 7 words, but it was taking people off the point. It was 8 distraction; 9 MR. OKRENT: I will read the sentence from Minogue's 10 memo, which I think Gillespie didn't give, which perhaps says 11 what Minogue is thinking most directly. "For the reason cited e 12 above, I believe a straightforward approach is to overtly \\ 13 adopt the prevention policy and use the numerical guidelines ~ 14 of core melt frequency and the safety goal as a means to this l 15 end." 16 MR. GILLESPIE: Yes. 17 MR. SIESS: It doesn't mention cost-benefit. 18 MR. OKRENT: No. He has been talking about 19 cost-benefit above, and he says that is a red herring in it. 20 MR. GILLESPIE: That happened to be one way it 21 looked to us, that the committee had proposed attacking that 4 f 22 problem. We are not hung up on the cost-benefit. We are hung l 23 up on the need for a phslosophy that says we are still going ( 24 to have the ability to g'o in on the prevention side of things, 25 and the traditional approach is to do that.

4 63 s 1 MR. KERR: What does not being hung up on 2 cost-benefit imply? Mr. Rowsome has just told me that it is 3 a very deeply embedded part of the Commission philosophy. He 4 didn't say hung up on it. 5 [ Laughter.] 6 MR. KERR: What does it mean when you say you are 7 not hung up on it? 8 MR. GILLESPIE: It means if conventional wisdom says 9 to do it, whether you have got some goal like a plant worth 10 fixed at $4 billion or $9 billion or $2 billion, if the 11 conventional wisdom says that it really does look like there 12 is a significant safety benefit to a particular plant from b \\ms 13 doing something, then we should get on with doing it. 14 ~ MR. WARD: And ignore the safety goal, in that case. 15 MR. GILLESPIE: You consider the safety goal, but 16 don't let it override your decision if the safety need exists; 17 that's right. There are two parallel processes in operation. 18 MR. S I E S S :. You take off the bottom, take off the 19 floor under the safety goal, leave the ceiling on, which is 20 essentially what the British have done. They go to as low as 1 21 practical once they get below their quantitative line. 22 MR. WARD: Now, when you said as practical, though, 23 he is not talking about practical They are saying if 24 judgment indicates that something ought to be fixed, they are 25 not going to worry about how much the cost

64 1 MR. SIESS: I would have to define practical without 2 bringing judgment in. It's either cost-benefit or judgment, it usually isn't one man, and 3 and sometimes, judgment 4 sometimes it is a collective judgment that has been arrived at 5 after six months of arguing with the utility or the industry. 6 So I think judgment covers it. You know, I don't like 7 conventional wisdom. Judgment isn't wisdom. 8 MR. OKRENT: Well, Mr. Chairman, I think we have had 9 a chance to hear from Mr. Gillespie, which would complete the 10 Staff views, and since you are behind your agenda, I would 11 propose we thank the Staff very much for helping to keep us 12 stimulated and confused by ourselves and go on with your next b \\ 13 topic for now. 14 We are supposed to come back. 15 MR. KERR: Would you be willing to accept a message 16 to Mr. Stello that we reafly do love him? 17 [ Laughter.] 18 MR. STELLO: Only if you give me a big hug and a 19 kiss. 20 MR. WARD: We'll have to do that during the break. 21 Let's take a 10-minute break. 22 [ Recess.] 23 MR. WARD: The next topic is the Diablo Canyon 24 seismic 10-year review, and I asked Dr. Siess to take over. 25 MR. SIESS: We don't call it the 10-year review, we

65 eS 1 call it the long term seismic program. This requirement for U 2 the reevaluation of the seismic design or seismic design basis 3 for Diablo Canyon was included in the license condition for to pick 4 Unit 1 chiefly, I believe, as a result of concerns 5 up on a comment we put in our letter of July 14th, 1978, which 6 was seven years ago almost, which recommended that the seismic 7 design of Diablo Canyon be reevaluated in about 10 years, 8 taking into account applicable new information, 9 At the time we wrote that, we thought it would have 10 been operating for a few years rather than a short time. The 11 license condition required this reevaluation of the seismic 12 design basis involving a number of things, starting with O) 's j 13 geology, seismo-tectonics, the magnitude of the earthquake, 14 the ground motion of the site and earthquake engineering. All 15 the aspects that we believed would affect the seismic design 16 basis, including the tau value, soil structure impaction, et 17 cetera. 18 The Licensee, Pacific Gas & Electric, came up with a 19 plan and this is what it looks like. It came out in January 20 on schedule. The Staff has undertaken a review of it. 21 Now the committee heard from the Licensee and their 22 consultants at a subcommittee meeting that was held around Los 23 Angeles on March 21 The subcommittee meeting was attended by 24 two members, Dave Okrent and me, and seven consultants, [v) 25 covering the total range of programs from geology on up

66 s 1 through the plant. 2 The Licensee presented to us at that time their 3 plan. Our consultants viewed it in general with favor. There 4 were some compliments made and no significant concerns 5 expressed. 6 The Staff at that time had just begun its review 7 which they have since completed, and our meeting yesterday was 8 to hear from the Staff their evaluation of the Licensee's plan 9 and to come up with our recommendations regarding it. 10 In the several months between the time the plan was 11 submitted and the time the Staff finished its review, there 12 were meetings with the Licensee and the Staff submitted a 13 rather long list of comments and questions and suggestions to 14 the Licensee, all of which were responded to, and as a result 15 of those exchanges, the Staff has concluded, I think fairly 16 straightforward 1y, that they find the program acceptable, that 17 they think it will do what the Licensee has been asked to do. 18 They have suggested that it be kept flexible so that 19 a change in emphasis is possible, so the Licensee is still 20 working on what he calls a scoping program to attempt to 21 narrow some of these, I believe most of the geologic aspects 22 of this down to a level that is manageable for the three-year 23 period. This must be completed in three years. It's a part () 24 of the licensing condition, and the clock starts running when 25 the Staff has issued its approval.

67 1 The Staff has issued a draft evaluation. They say O 2 it is not final until they have heard from the ACRS, at which 3 time they will either agree with us or disagree with us. I l 4 think those are the only two alternatives. or first of all, the Staff 5 Now the subcommittee 6 has consulting help. They have a geologist, Dr. Slemmons. 7 They have the U.S. Geological Survey. Our consultants, 8 geological consultants, were particularly pleased that 9 Dr. Slemmons will be working with the Staff. They have a 10 great deal of confidence in him. 11 All the rest of the geologists and seismologists are 12 working for the Licensee, and I think between the Staff and 13 the Licensee, there is not a California seismologist left, is 14 there? Although I think Dr. Okrent suggested one that might 15 be considered. 16 MR. LEWIS: We breed them in California, you know. 17 MR. SIESS: Yeah, I know. 18 Now, this is a complex program and what I would like 19 to do is simply tell you that with one exception, as far as 20 the program is concerned, the Staff is satisfied that it is a 21 workable program, that it will achieve the objectives. 22 Of course, this is not the end of it for the Staff. 23 They will have continuous interaction with the Licensee. (v\\ 24 Progress reports at six-month intervals, or reports 25 tt six-month intervals

68 i MR. BROCOUM: Progress reports are quarterly. This 2 is Steve Brocoum, Geosciences Branch. And meetings at a 3 minimum every six months. 4 MR. SIESS: Minimum of six months. So the Staff has 5 found this acceptable, 6 The subcommittee and consultants in March found it 7 acceptable. The subcommittee in its meeting yesterday -- and 8 that was Jesse Ebersole, Carson Mark, Dave Okrent and me -- f 9 found the program acceptable, with one exception, and the 10 exception is one taken by Dave Okrent to the proposal by the 11 Licensee in satisfaction of the requirement that says POSE 12 shall assess the significance of the conclusions drawn from N 13 the seismic reevaluation studies utilising a probabilistic 14 risk analysis and deterministics studies as necessary to 15 assure adequacy of seismic margins. 16 And this has been interpreted in various ways, but 17 one possible way of looking at it is that if it turned out 18 from all the geologic, seismologic and other studies.that the 19 original seismic design bases were adequate, those to which 20 the plant was redesigned, then there would really be no reason 21 to assess the significance of the conclusions, except to say 22 that they confirmed previous findings, and that nothing would 23 be necessary, /" 24 The Licensee has not been that optimistic and has r 25 proposed making a PRA beginning essentially now, having it i

69 1 ready to use in the assessment of the significance. He has k 2 proposed not just a seismic PRA, but a full scope PRA, taking 3 into account internal as well as external initiators which we 4 think is useful 5 But he has proposed only what has been called a 6 Level 1 PRA. 7 Now as I understand it, Level 1 carries through core the core melt 8 melt scenarios, develops a core melt 9 frequencies, and what is referred to as plant damage states 10 corresponding to each of those scenarios, whether it's an 11 early or late or whatever, and somebody can explain that in 12 more detail, O 13 Level 2, I believe, goes to source term within 14 containment. 15 MR. OKRENT: I think we had better have a better 16 definition of Level 1 and' Level 2. 17 MR. SIESS: Well, I am reading this from something 18 that the Staff gave us at the subcommittee meeting. 19 MR. ISRAEL: I will go into it later. 20 MR. SIESS: Well, what I have seems strange because 21 it says Level 4 goes to consequences, and that sort of left 22 Level 3 in limbo. But Sandy will explain that to you. 23 The issue is simply this: 24 The Staff is satisfied with a Level 1 PRA as a basis ) 25 for assessing the significance of whatever conclusions may

70 1 come out regarding the groundmotions. The subcommittee at the fmgU or at least the two people that were there -- 2 March meeting 3 suggested that there would be some advantage to the Licensee 4 in going to consequences'and not stopping with core melt, and 5 the Staff said well, if they go through Level 1, we can 6 extrapolate that to the consequences. 7 This was discussed at length at the meeting 8 yesterday. In fact, it was almost the only subject of 9 discussion. The three members of the subcommittee other than 10 Dave agreed or decided they would agree with the Staff that 11 Level 1 would be adequate. 12 Dave believes that they should go to Level 2 or 3, 13 depending on what the proper definition is. Dave would like 14 to see them go to consequences. 15 MR. OKRENT: No. 16 MR. SIESS: Well, okay, Dave will have a chance to 17 say and we will figure out how to go about it. 18 So what I am proposing is that you accept the 19 recommendations of the four subcommittee members as far as the 20 program plan is concerned on matters other than the PRA level, 21 and that you hear the story from Dave, his reasons, the 22 Staff's reasons, the Licensee's, if you wish, why they think 23 Level 1 is adequate, and then let the Full Committee decide ( 24 whether they want to accept this, the Level 1 PRA, or want to 25 make a recommendation for the higher level. That is the issue

71 (~5 1 I would like to put before the committee. V) 2 First I am asking you to accept our recommendation 3 on the other aspects of the reassessment of the earthquake 4 effect, and to hear and decide on the issue of what kind of 5 PRA is needed in order to assess the significance of this. 6 Now the other members of the subcommittee, Jesse and 7 Carson, have I spoken for you? Or you can speak for 8 yourselves. 9 MR. EBERSOLE: I just thought I would mention that 10 Level 1, as I understood it, did include consideration of 11 containment damage as an initiator to core damage. 12 MR. SIESS: Oh, yes. 13 MR. EBERSOLE: And that's as far as it went. 14 MR. MARK: Well, I don't think I have anything to 15 add to your account, Chet. There are, of course, other 16 reasons which might appeal to the Applicant to make a more 17 extended PRA. There'is a very good chance of learning 18 something, and being able to explain things in connection with 19 a plant for future arguments. 20 I guess a PRA has never been done on Diablo Canyon, 21 full scope. 22 MR. SIESS: Full scope. We did a partial seismic 23 PRA -- well, a risk analysis, seismic risk many, many years 24 ago, but not really what we are talking about now. 25 MR. MARK: Well, there are a number of things which l

72 1 I'm sure he has in mind more clearly than I, like the severe 2 aooident policy and all kinds of things where a PRA might be 3 handy to have in your briefcase or in your trailer truok, 4 whichever they go into. 5 MR. SIESS: Dave, which would you think would be 6 best? To let the Staff say why they think Level 1 is enough 7 and then you give your arguments? 8 MR. OKRENT: Well, I think it would be well to have 9 someone define in the first place what Level what the Staff 10 thinks is adequate and 11 MR. SIESS: What it is and why they think it's 12 adequate? \\s-13 MR. OKRENT: Yes, what it is and why they think it's 14 adequate, and then they can define Level 2 as they interpret 15 it and at that point we could have a discussion, because right 16 now we don't have it. 17 I will wait until their 18 MR. ISRAEL: Let me briefly go over the analysis 19 process. Most of you are familiar with it, but at least to 20 put everything into perspective. PRA analysis usually starts 21 off with what we usually call the front end, where the plant 22 systems are analyzed in an event tree, fault tree fashion to 23 determine those sequences that will lead to core melt. ( 24 The sequences may have various initiators and what 25 you are concerned with are failures of various systems so that >Y

73 1 in effect you would end up without core cooling and you s 2 would end up with a core melt. 3 In the past this analysis has resulted in 4 determining various plant damage states. Plant damage state 5 in effect tells you something about the core condition, core 6 melt and the containment condition. The one of particular 7 interest in terms of the seismic analysis would be containment 8 failure prior to core melt which would occur if, in fast, the 9 earthquake failed the containment which would then obviously or maybe not so obviously,-- fail the piping running into 10 11 the reactor and core melt would ensure from there. 12 You have other plant damage states that could 13 occur. One is core melt without any containment melting, and 14 this usually occurs because you lose all electric power in the 15 plants, both offsite and onsite. This can occur because of 16 the seismio or it could occur for other reasons as well. 17 There are other plant damage states that occur. 18 Event V, which you have all heard about, which may not be 19 specific to seismic, but in this situation we have containment 20 bypass prior to core melt. Event V is failure of the low 21 pressure ECCS piping outside of containment so that you have a 22 path outside of containment, with potentially subsequent oore 23 melt. () 24 There other damage states dealing with transients 25 and LOCAs, which basically are early core melts if you had a

74 /~N 1 large LOCA and didn't have injection, or you could have later 2 core melts because of small breaks or transient situations 3 where you lose core cooling and ultimately the core would 4 fail. And the concern here is the availability of containment 5 cooling, and this would be determined by the analysis of the 6 plant systems in the fault tree process. 7 This is basically what Level 1 is. It is 8 essentially a plant damage state analysis. That is the front 9 end. The back end, which will differentiate between Level 2 10 and Level 3, the back end tries to characterize or 11 characterises what happens to the containment given any one of 12 these plant damage states. \\ 13 Let's take, for instance, core melt without any 14 containment cooling as a plant damage state. You would be 15 concerned about the split fractions as to where the potential 16 containment failures would occur, how they would occur. You 17 have heard about steam explosions, the alpha mode where the l 18 vessel would explode and go through the containment. 19 You have heard about early hydrogen burns that could 20 fail the containment. There are late hydrogen burns that l 21 could fail the containment. There are overpressurization 22 failures of the containment as the pressure just builds up and 23 there is an overpressure of the containment. Or you could go 24 to basemat melt-through or no failure at all. 25 So all of those are different split fractions dealin f i m- ,7 -.~.w,., _ _ _y

75 m 1 with a given plant damage state. So for each one of the plant 2 damage states, I could then go into various potential 3 containment failure modes or no failure at all. 4 Those split fractions are usually determined under 5 Level 2. Along with each one of those split fractions, which 6 are usually characterised by a release category, if you will, 7 the analysis then calculates a source term for each one of 8 those release fractions. 9 So at the end of Level 2, I now have plant damage 10 states which I now multiply by various failure mode, 11 containment failure mode split fractions, and for each one of 12 those I have a source term. I can now go over into Level 3, l 13 which is the consequence analysis, and the consequence 14 analysis takes each one of those source terms for each one of 15 those containment failure modes and determines the offsite 16 consequences. 17 Our past experience in terms of offsite consequences 18 has been that early fatalities are dominated by containment i 19 failure prior to core melt. I say this, and that is based on 20 the spread of plant damage state frequencies that we have seen 21 in the past. Obviously, if we had some b*nign plant damage 22 state that suddenly had a frequency of 1, potentially that may 23 take precedence. 24 But based on the spread of plant damage state i 25 frequencies that we have seen in the past in the various

f g .m 76 g 1 plants we have analysed, early fatalities are dominated by ) l melt," and those types of 2 ' containment failure prior to core j j 3 sequences are basically the one that you could potentially get j 4 with a seismic event where the earthquak --vocks the 5 containment away from the auxiliary b u i l'd i ng or tails the i 6 cont a i nmen t' s omehow. It could also be Event V. 7 You also potentially could get early fatalities from 8 long-term containment failure, and the way we arrive at that 9 is usually through the station blackout sequence or what I 10 will call core melt without any containment cooling, so that 11 over a period of time the pressure builds up and the 12 containment fails and that is released to the environment. Ok,) 13 You could get early fataliti s from that. s 14 There are other plant damage states, but those that 15 seem to have c o n t a i n m e n t' c o o l i n g, d o not seem to contribute to p 16 early fatalities, i j e r pe[ son 17 In term 4foi $atent fatalities and rem, the 18 dominant contributor appears to be, from past experience, core 19 melt without containment. cooling, again, which also looks like 20 an extended station blackout type of sequence. 21 There are several reasons why the Staff didn'lt 22 require the Licensee to extend his analysis beyond Level 1, 23 which would he the plaMt damage state levet. The foremost 24 reason, and this goes back a year ago and we are still in this g 25 mode,_.was that we are in this source term upheaval There is 9 4

\\ 77 /'N 1 the old source term and the new source term, and we are in the 2 middle of the breach here and this study is not going to be 3 done for three years, and I saw now way how to indicate to the 4 Licensee what he should be doing with the source term. 5 For that reason, that was one of the primary reasons 6 for not requiring him to do the back end. 7 Let me tell you that in terms of the source term 8 effort, evidently there will be a Staff report coming out 9 within the next six months, I believe, or maybe in the next 10 couple months, NUREG-0956, dealing with wherever the Staff 11 thinks it is on the source term. I think within the next year 12 .or year and a half, the Staff will also come out with a sl 13 report called NUREG-1150, which will be a mini-update, I 14 think, of WASH-1400 in terms of source term, where they will 15 analyze five surrogate plants: Surrey, Zion, Sequoyah, and a 16 couple boilers. 17 MR. EBERSOLE: May I ask a question, Sandy? 18 MR. ISRAEL: Yes. 19 MR. EBERSOLE: How do you interrelate Event V to the 20 seismic event? If you just 21 MR. ISRAEL: Potentially the seismic event could 22 take out that pipe as well, 23 MR. EBERSOLE: But if it's going to take that out, 24 then it's not merely Event V, it's a host of other similar 25 events. For instance, reactor water cleanup system. There

78 i 7-~g 1 is an event similar to V. Or the one we were just discussing 2 earlier, the main steam line failures to HPSI systems, which 3 doesn't apply to Diablo. 4 MR. ISRAEL: What you pointed out was that I 5 characterized that Event V obviously comes from internal 6 events, but potentially 7 MR. EBERSOLE: There are quite a few Event Vs. l l l 8 MR. ISRAEL: Yes. If you fail the containment, you 9 fail the whole thing. But you pointed out a good point. The 10 seismic calculations they are doing for Level 1 would then 11 give frequency for this containment failure prior to core 12 melt, which is probably going to be a dominant factor in terms O k-13 of early fatalities, and the most likely type of sequence for m 14 seismic is the -- well, the one of significance is station 15 blackout because a large seismic event takes out offsite 16 power, and once you start getting into the electrical area, it 17 will wipe out a lot of other things. 18 Those are probably the two sequences of concern 19 coming out of the study, and those would be identified in 20 plant damage state. l l 21 But anyway, to get back, within about a year and a 22 half, we should have this five-plant study using new source 23 term. The new source term work includes considerations of new 24 containment or whatever the containment failure modes are on 25 the table at the time of doing the work, and there are some

,u .L 79 1 g'~s 1 new ones since WASH-1400. Also, it would come up with 2 containment failure split fractions: you know, what fraction 3 would go into the various bins such as hydrogen burn failure 4 or long-term overpressurization, what have you, and they would 5 also calculate source terms for each one of these based on 6 whatever the technology is that they are into. And presumably 7 the Staff would be embracing this sort of thing. 8 So this, then, gets to one of the next concerns as 9 to why we didn't require a level -- anything more than a 10 Level 1. Ordinarily when people start doing Level 2, Level 3 11 analyses, they are very expensive. There is a lot of computer 12 calculation, a lot of phenomenological calculations in terms 13 of containment loading, sensitivity studies dealing with oore 14 debris coolability and core / concrete interactions, et cetera. 15 These can become very expensive, and even our review 16 of these becomes expensive. As I mentioned yesterday, we 17 spent over $500,000 reviewing the Indian Point work several J 18 years ago. 19 There is an alternate and less expensive way of 20 estimating offsite consequences, and here again, I want to i 21 stress that these are estimates, and I also want to stress 22 that the whole purpose of this Diablo Canyon study is to 23 investigate seismic characteristics of the site as opposed to O 24 some of the other PRA work which we have done which has been g' i 25 more geared to looking at containment potentials. i

80 1 Certainly we have today the siting study work that 2 was done at Sandia several years ago, which looked at every 3 site in the country and came up with surrogate source terms. 4 They also looked at Diablo Canyon, and people who are familiar 5 with the back end analysis with a little clever extrapolation 6 and interpolation of this information could come up with 7 estimates of what the consequences would be at Diablo Canyon 8 using whatever the front end core damage frequencies are, but 9 would be based on old source terms, 10 I suspect that three years hence what we really 11 will have, and another shortcut methodology, we will have this 12 five-plant study, which includes Surrey and Zion, which ) '~ / 13 probably encompasses the internal workings of the containment 14 concerns, and that will provide source terms and containment 15 failure split fractions, and I suspect that what we can do 16 then is we will use that along with the Diablo Canyon plant 17 damage states to come up with consequences, 18 MR. EBERSOLE: I wonder if you would explicitly tell 19 me what you mean by split fraction. 20 MR. ISRAEL: Okay. Some people call them split 21 fractions and sometimes they call them conditional 22 probabilities. For instance, the alpha mode, which is the 23 steam explosion inside the pressure vessel blowing the head ( 24 off, et cetera, has a conditional probability of like 10 to 25 the minus 4 that the containment will fail in that mode.

81 i e 1 For a plant that has a core melt without any 2 containment cooling, the conditional probability for long-term 3 overpressurization failure may be like .4 or .5, and the 4 probability that basemat melt-through may be around .3, .4 for 5 that, and maybe 1 there would be no failure at all. So 6 really the conditional probabilities of this plant damage, 7 this scenario, which is in some plant damage state going into 8 some sort of containment failure mode. 9 So the thing I was trying to stress is that we have 10 an alternate way that is not detailed and less expensive-for of what 11 coming up with estimates -- and I stress estimates 12 the offsite consequence is. ) \\d 13 I suspect that at the time that the Disblo Canyon 14 study is completed, depending what we see in terms of the 15 plant damage state frequency may push us in one direction or 16 another in terms of how much scratching we do in the 17 consequence area. 18 A third reason that we didn't push them into doing 19 back end analysis is that traditionally the Staff has 20 performed these back end analyses for PRAs that have come i n,, 21 and even though the licensees or the. applicants have submitted 22 back end analysis, we seem to do these audit calculations, and 23 it is like paying twice for the same amount of work, in 24 addition to which, at the end of all of this study, the 25 licensee will come in and say, hey, this i s what my plant i .-y m m

i 82 1 looks like, and the plant damage states will obviously 2 identify weaknesses in terms of the seismic situation and it 3 may suggest all by itself potential fixes. 4 At Indian Point there were buildings being 5 together. I think at Zion there was concern about loss of the 6 service water pumps. I think Indian Point 3 had trouble with 7 the fuel oil line going to the diesel generators. 8 The Staff -- you heard earlier Frank Rowsome talking 9 about the fact that the Staff is responsible for coming up 10 with a cost-benefit analysis if, in fact, we want to backfit 11 the plant. They say, hey, we looked at your results and wo 12 think that you ought to do X, Y, Z, and therefore we are on \\%./ 13 the hook for doing a cost-benefit analysis that would require 14 us to 1...orm an offsite consequence analysis. 15 The other feature that struck me was that this 16 happens to be a low population site. There are approximately 17 200,000 people, roughly, within about 50 miles of the plant. 18 The average site in the United States is about a million. 19 Indian Point had about 17 million. So just to give you some 20 sort of perspective on the potential societal risks that we 21 are talking about. 22 There is another reason that I have that I think is 23 quite valid, that I stress that the whole thrust of this ( 24 Diablo Canyon study is dealing with the seismicity and geology 25 and what have you of the site, and it really isn't geared to

83 1 examining -- it's not good form for resolving the ongoing N 2 source term severe accident work. 3 I am somewhat concerned about that because, as I 4 mentioned yesterday, the Indian Point PRA that came in had 12 5 volumes. Three volumes dealt with core melt from internal 6 events, one volume dealt with core melt from external events, 7 one volume dealt with a summary of the whole shebang, and 8 seven volumes dealt with containment analysis, source term 9 analysis and consequence analysis. Somehow I think that is 10 the wrong emphasis for the effort at hand. 11 Basically the summation of all those were the 12 reasons that the Staff felt that it was acceptable to just 13 stay with the Level 1. 14 MR. SIESS: Thank you. I am pretty sure those are 15 the same reasons we heard yesterday that the three of us found 16 compelling. 17 I would suggest that we hear Dave before we 18 question. 19 MR. OKRENT: I must say I am not sure that you found 20 them compelling, plus, in fact, I'm not sure you had a clear 21 picture of the difference between what is required at Level 1 22 and Level 2, but I will just leave that as an aside. 23 MR. SIESS: I didn't change my opinion. 24 MR. OKRENT: First, let's look at the difference ) ( 25 between Level 1 and Level 2. And I am not talking about Level

84 /~ 1 3. (NI v' 2 For Level 1, as they say, they would calculate the 3 accident states. They would not have evaluated containment 4 . failure modes, except those that resulted from the earthquake 5 itself. They would not have evaluated the containment 6 capability to withstand pressure and temperature, nor would 7 they have evaluated whether this design basis, I will call it, 8 or as-designed capability to withstand pressure and 9 temperature in any way is weakened by a severe earthquake, 10 because it's going to be, in their study, either the 11 containment fails -- and I don't believe for a minute that a or some penetration fails or 12 building is going to fail 13 not So, in fact, they're going to have a big hole in their 14 ability to assess what I would call -- what's called the 15 likelihood of different release categories. 16 Obviously, the containment capability is vital in 17 this. i 18 Secondly, Diablo Canyon has different configurations 19 than Indian Point and Zion and Surrey with regard to 20 containment cooling capabi'.ity, with regard to AC power, with 21 regard to service water, for example, and one particular way 22 in which it will be different from these is that since, in 23 fact, we are particularly assessing the effect of seismic, the 24 review presumably will consider a~ series of more complex g l 25 transients than is initiated by the different kinds of l

85 i failures or combinations of failures that you ordinarily just \\ 2 don't look at in an ordinary PRA, because you say that many 3 things won't occur randomly, whereas you can lose all of your 4 alarms and much of your control room instrumentation and a few 5 steamlines, et cetera, et cetera, all with the one earthquake 6 here. And it is, I believe, going to be a different problem 7 to try to evaluate the operators' ability, for example, to 8 restore a lost diesel generator or a lost cooling pump or so 9 forth, as compared to what is done in the -- that you cannot 10 just translate from the other PRAs. 11 So I think the Staff will be in a poor position to 12 make estimates of release -- of the mode of containment O ( 13 failure and the frequency of containment failure in release 14 categories. 15 Furthermore, I don't believe for a minute it should 16 be the Staff that's making that estimate. I think the 17 Applicant, the Licensee, in the first place, is the one who 18 has better knowledge of the plant and so forth and will have 19 done the PRA work that would help you somewhat in doing the 20 Level 2 work, and it doesn't make sense to me for the Staff to it would be 21 do it. And I think what they do would be very 22 suspect, because of their lack of the kinds of things I've 23 just indicated. b 24 I think the basis given by the Staff as the primary 25 basis for choosing Level 1 is just -- just doesn't hold

i 86 -1 water. He said the source term upheaval is the one for only N_/ j 2 going to Level 1. I must say, we have been getting Level'3 3 PRAs from a lot of Licensees in the last few years. 4 MR. KERR: Level 3 or 2? 5 MR. OKRENT: We've been getting Level 3. That means-l 1 6 they not only do -- I meant Level 3. In other words, if you t 7 look at Limerick and Shoreham and Zion and Indian Point and and I haven't seen Susquehanna -- Midland -- they 8 Millstone l 9 been doing Level 3, which means they had to do Level 2, of 4 10 course. I'm not suggesting in any way that Level 3 be done, 11 where there's a lot of calculation of consequences, a lot of 12 these volumes that he's been talking about, a lot of the 13 computer time that you've been tal?ing about, but I do think s i 14 it's relevant to understand the behavior of the containment, f 15 given a coremelt, and how it affects containment failure 16 likelihood. 17 On the one hand, to say the source term upheaval 18 leads us to say, "Only go to Level 1," but on the other hand, 4 I 19 we're going to have a method of estimating all this shorthand 20 again seems to me to be somewhat contradictory. 21 With regard to the expense of Level 2 added to Level 22 1, I think what we heard yesterday -- I can't remember the but if the total cost of this entire study is 23 exact number I 24 eight to ten million, I vaguely recall, and the cost of the ( and correct me, if ) 25 PRA, I assume, is two or three million I

87 or higher, four million, I think going from a /~ 1 I'm wrong (% 2 Level 1 to a Level 2 will -- let's say it is three million -- 3 will not increase it by more than 25 percent, I think, and my the PRA. 4 guess is that it would be less than 25 percent 5 Maybe I'm wrong, but I don't think so, because they have the 6 advantage of the work done by Bowsey & Associates and so forth 7 for Zion and Indian Point, and the other utilities have used 8 this in their PRAs. 9 And just what additional new knowledge is developed 10 between then and whatever is the time when they start trying 11 to evaluate the exact build-up of pressure and so forth is 12 .they are not going to be running experiments or developing new 13 codes or anything like this, and the ground was broken, 14 really, at Zion and Indian Point. 15 So for a variety of reasons, I think it is the 16 Staff's position, it is the wrong one. I think, I'm sorry to 17 say, I think my fellow subcommittee members maybe didn't 18 appreciate the points I was trying to make 19 [ Laughter.3 20 MR. OKRENT: But I think the committee, in fact, 21 should recommend a Level 2. 22 MR. SIESS: Dave, hel; me on something. In the 23 session preceding this one and in some other discussions, I've ) 24 gotten the impression that you thought a coremelt guideline 25 was a very important, almost overpowering safety goal. And if . ~

88 1 that impression is at all correct, why isn't it good enough 2 for evaluating the seismic design basis at Diablo? 3 MR. OKRENT: Wait. In the first place, I have in i 4 writing made the point to the Staff that a coremelt guideline 5 is not adequate and that when they talk about meeting a 6 ceremelt guideline without looking at the risks, that is an 7 incomplete look. So I won't accept the original attribution, 8 okay? 9 I think it's relevant to know the coremelt 10 frequency. 11 MR. SIESS: The health effect guidelines are more I 12 important, then. 13 MR. OKRENT: No. I have in general and I'm not I have in general said, if we 14 changing my position here now i 15 could get an estimate of containment release frequency, we 16 have the essential information for estimating the safety of 17 the reactor, and it's the same position I had a year ago. 4 18 MR. SIESS: Incidentally, I might point out that the 4 19 additional cost i s not negligible, but'it is not a prohibitive 20 cost by any means. It might have an impact on the time. 3 l 21 MR. OKRENT: It will have an impact on the time, if 22 it is decided three years from now that it is needed. If it 23 is decided early on, it can be -- much of the work can be done 24 concurrently. 25 MR. SIESS: Unless the source term changes in three 1 -=~s w, -,,-.,,.--.,_g-.7,

89 4 1 years. v 2 MR. OXRENT: No. But again, they would make some 3 estimate of source term with uncertainties. I know how PLG 4 does this. And if there is new information, they would either 5 change their distribution and come up with new numbers or 6 whatever. The methodology is set up, and they can use new 7 information on a particular isotope or whatever they are 1 8 dealing with. 9 MR. SIESS: Mr. Chairman, I am going to turn it back 10 to you. We have here from PG&E a representation that they 11 will answer questions. I don't think they have anything they 12 want to say. They have proposed the Level 1. The Staff has O s 13 accepted a Level 1 And they answered some questions 14 yesterday about possible costs and impact on time. 15 They have indicated in response to the Staff that i 16 depending on how the results come out, after they get the new 17 seismic design basis and look at the Level 1 PRA, they may 18 want to go on beyond that, as well you might, because the 19 farther out in the consequences you go, I think the better 20 this plant is going to look. But that's just my guess. 21 So what I'd like to do is turn it back to the 22 Chairman to get the sense of the full committee. I'm in a 23 position to draft a letter which says we approve the program ) 24 and comment that the proposed Level 1, the Staff has accepted 25 it, and we agree, to which they can add additional comments or

90 1 anybody else. Or I am prepared to write a letter that says we G 2 approve the program, and we think that the PRA ought to be .3 beyond Level 1, to Level 2 or Level 3 or whatever we might 4 say, and I propose that those are the alternatives, and I i 5 think if you could find out how the committee wants to go, 6 even now or Saturday. 7 MR. WARD: All right. Thank you, Bill. 8 MR. KERR: What would you anticipate one would do 9 with the results of the Level 2 PRA that one could not do with 10 the results of a Level 1? 11 MR. OKRENT: Well, the Level 1 doesn't -- 12 MR. KERR: I understand the difference between the s 13 two, I think. 14 MR. OKRENT: It doesn't give you a good handle on 15 the releases and 16 MR. KERR: But suppose now we have a good handle on 17 the releases. What does one then do? This is an operating 18

plant, i

19 MR. OKRENT: Yes. But they might 20 MR. KERR: I'm not trying to be critical I'm 21 trying to understand. 22 MR. OKRENT: Well, they may have a coremelt 23 frequency like Indian Point, which the Staff estimates would 24 he about 4 times 10 to the -4. And if you didn't have an g 25 evaluation of containment capability, that could look quite

~. 91 .1 threatening or more threatening than it is, but you can't just -w V 2 say that containment is equivalent to Indian Point. In fact, 3 the Staff argues that Indian Point has a better than average 4 one. 5 MR. KERR: I did not sharpen the question enough. 6 One consequence of carrying this out would be that you would 7 decide to shut down the plant. I sort of went to an extreme. 8 Now short of that, what would one anticipate that one might i 9 do? Rebuild the containment, for example? there might be some procedure 10 MR. WARD: Could I 11 changes or some instrumentation or something that would be 12 MR. OKRENT: Well, penetration. There may be one i x_/ 13 penetration that will go only to five psi above the pressure 14 at which it was tested. And if they don't look for 15 containment pressure capability, you don't know that. 16 Now you know at*Sequoyah, they fixed up one major 17 penetration, for example, and that may not be the only reactor 18 that did it after looking at containment pressure capability. 19 So I really can't understand, in fact, spending eight to ten 20 million dollars, of which three or so is on this FRA, and not 21 spending this additional whatever it is, 300 to 500 K, to get 22 to Level 2, and I'll tell you, as I say, it's stopping at the 23 wrong point. ) 24 MR. WARD: Could I make a comment? I have to leave 25 for a few minutes.

92 1 I guess from what I have heard, I am very much in you know, my personal 2 sympathy with what Dave is saying 3 position. It seems to me, the point of doing the seismic PRA 4 is to look for what might be some new relationships, I mean, 5 that are unique to the sort of common failures you can get 6 from a major seismic event that could effect, you know, both 7 the ability to cool the core and the ability for the 8 containment to perform as designed. And it just seems to me 9 that if the PRA isn't probing for those things, it isn't of an 10 awful lot of use. 11 MR. SHEWMON: It is certainly probing for core i 12 damage or coremelt, isn't it? 13 MR. WARD: It is, but that's just half the story. 14 MR. SHEWMON: I know, but that's not quite the way 15 you worded your statement. So you are onif concerned about 16 what you feel is left out, is the performance of the 17 containment. 18 MR, WARD: Yes. And I know they are looking I think what I heard at 19 specifically at, I guess 20 containment failures from seismic. But it strikes me that 21 there are some interrelationships which the PRA is the tool to 1 22 use to examine those. 23 MR. SHEWMON: Now the reason for this whole exercise 24 was that there is a possibility of higher seismic loading here 25 than other places, and therefore they are supposed to go back

m. _ _ _ =. - _ _ _ _ _. _. - _. - - ~ 93 1 and look harder. Is that sort of a price for the license? 2 MR. OKRENT: If one wants to be a little bit 3 lengthier, when the committee first reviewed the operation in 4 1977 or '78, we agreed with the design basis, but we didn't 5 agree with the methodology that the Staff was proposing, 6 whereby they arrived at a design basis. And at that time, we 7 suggested, look, we are still in a learning stage with regard 4 ] 8 to earthquakes and in particular earthquakes around here and 9 so forth. It would be good to take another look at the 1 77 10 seismio design basis in ten years; that's what we said in 11 or '78, t 12 And then whenever we were reviewing something, 13 quality assurance or something, we reminded the Commission 14 that we had made this recommendation, and the Commission 15 picked it up, not when we first made it, but when they were 16 reviewing the quality issues, okay. I 17 MR. MARK: Well, there was another point, as I 18 recall it. When we wrote that first letter, there was a 19 really wide spread of opinion as to just what the seismic 7 20 setting was at Diablo Canyon, and it was also known that there 21 was a lot of further exploration going to occur in the next 22 few years. And while it looked all right on the basis of what 23 we sort of guessed was the case, it was possible that it 24 should be looked at again. 25 MR. SHEWMON: That is a major part of this program. i 'I a nn ,e ,,~,------n n ..,e..,,,,.----n.,,-----.--------n-----,~,,--------.n--- ., - - -. -., ~ ~ ~w---.

94 j'^g 1 MR. OKRENT: Yes. 2 MR. EBERSOLE: Can I ask a question? Dave, maybe 3 I'm just beginning to see the light. Are you saying that what 4 you are bothered by is that even if the containment damage 5 itself was not an initiator, you could have a damage state, J 6 some other kind of damage state, that led to oore damage, but 7 you would have a containment which wouldn't work? 8 MR. OKRENT: Well, in what they are proposing to do 9 at Level 1, they won't evaluate the containment capability, i i 10 even its pressure capability. 11 MR. EBERSOLE: But it could be damaged by the 12 earthquake, not having 13 MR, OKRENT: Furthermore, it could be damaged but 14 not fail, and I don't know how one would get at that. But 15 again, they wouldn't look at that, since they are not going to-16 look at the containment. 17 MR. EBERSOLE: But the containment can be damaged 18 like anything else, and then you can have a damage state, but 19 it will not contain. if they can 20 MR. OKRENT: Well, if the containment l 21 predict a loss of containment integrity due to the earthquake, 22 that will be done in Level 1. 23 But degradation of ability to withstand pressure, I 24 don't know. Not in this. They are going to have to look at ( 25 pressure capability. ... - ~.

95 fs 1 MR. EBERSOLE: Well, I thought that's what ? J 'ud 2 containment failure would be, a loss of ability to do anything 3 it's supposed to do. 4 MR. OKRENT: Well, Jesse, you are the man who 5 frequently says things aren't always on or off, or black or 6 white; right? 7 MR. EBERSOLE: Yep. 8 MR. OKRENT: Okay. Well 9 MR. SHEWMON: To what extent are we telling them to and I am using this only as a 10 go up for common mode 11 paraphrase, you know, bring me a rock. 12 MR. OKRENT: No, no. A/ 13 MR. SHEWMON: But you are talking about there is 14 some kind of damage, we're not sure what kind of damage, and 15 there is now an interaction between this damage which didn't 16 cause failure in some future event, which may cause leakage. 17 MR. OKRENT: No, I'm sorry, but there are some 18 things mixed up in here. If you do a PRA, you want to 19 estimate the releases from the containment. You have to 20 evaluate the capability of the containment to withstand 21 pressure and temperature. You do that in a Level 2, not in a 22 Level 1 23 You furthermore have to evaluate them, the build-up 24 of pressure with time, to estimate, so you look at the cooling 25 capability and if it is lost, will it be restored and so forth

96 1 in time before the pressure exceeds whatever is the failure O 2 point and so forth. 3 Again, that would be part of a Level 2 examination. 4 There may be some special things in a severe earthquake that 5 have not been looked at, That's what Dave Ward was suggesting 6 a little bit that they would also look at. 7 I suggested one thing, namely, there is going to be 8 a lot more room for confusion in the control room since they 9 will have lost much of their instrumentation automatically, 10 and there may be much more complicated transients, 11 furthermore, because you have multiple failures of nonsafety 12 systems. 13 MR. 81ESS: But that's still covered in Level 1. 14 MR. OKRENT: Well, not ordinarily, and how they will 15 do it, I don't know. 16 MR. SIESS: But*you don't have 17 MR. OKRENT: The ability of the operator to recover 18 from these, you don't have to do unless you want to find out, 19 you know, how long does this transpire, and when does it get 20 water back to the containment sprays, if that's the cooling 21 mode, and so on. 22 So for a Level 2 you have to examine the ability of 23 the operator to recover under these circumstances, to fix the ( 24 diesel generator under these otroumstances or whatever. 25 Whereas if what you have to do is the damage state, you can

97 f"N 1 stop at an earlier point, you don't have to get to the test of 2 whether the containment has failed or not. 3 MR. SIESS: Let me try to make a point. The basic 4 objection of this reevaluation was to look at the seismio 5 design basis which is a spectrum input at the base of the 6 structure. It's arrived at by looking at what happens at the 7 site, looking at the soil structure interaction, et cetera, 8 and you find out what happens at the site by starting 9 somewhere out with an earthquake and transmitting it in and so 10 forth. 11 The first result of this thing, the first three i 12 items on the list, will give them a new design basis or a 13 spectrum of design bases, maybe with probabilities on them. 14 If those design bases are different, then there needs to be 15 some way of determining what is the significance of that 16 difference, and this fourth requirement was put in that once 17 they get the results of this, they assess the significanoe of 1 18 it. 19 If it is half the previous value, I don't even see 1 20 why you would have to do a PRA. If it is greater than that, 21 you certainly want to assess the signittoanoe to see if you 22 have to shut the plant down, because it's now somewhat larger 23 design basis. 24 I feel confident that if it comes out significantly 25 different and the assessment based on a Level 1 shows there is _ - - - _-_- _ __-__~_

1 l i 98 1 significance, nobody is going to have to tell them to go to a 2 Level 2 or 3 or certainly we're not going to have to tell 3 them. I think somebody is going to look at that because the 4 low population and a number of other factors are going to be 5 in their favor, that the significance, I think, is going to be 6 smaller, the farther down the line they go in the PRA. 7 Now that's a judgment. So I would rather see them 8 get on with this job in three years, come up with a revised 9 design basis, and then look at the significance. And I am 10 convinced that if it looks significant, we are going to find 11 out about it and something will be done about it. 12 Now they have to get shut down for a year or two 13 while they go to that Level 2 or 3. That's another story. 14 incidentally, Paul, this is an insurance policy in 15 one respect because as new geological evidence is developed 10 16 years from now, what they are doing will be in such a form 17 that they could factor that new evidence in very easily and 18 determine its effect. 19 So I think it is a very good insurance policy 20 against future improvements and knowledge which are 21 inevitable. 22 MR. EBERSOLE: Do new geological findings always 23 have a tendonoy to make things worse? 24 MR. SIESS: New geological findings tend to make ,v 25 worse, but the assumption here to start off with on the Hosgri

99 1 was so far in one direction that I think there is at least a "N \\_/ 2 decent chance that it will end up less. 3 Then they are going to do a lot more in terms of 4 getting from the earthquake into the site, it's more refined 5 than what they did before. + 6 I think our consultants felt that there was about a 7 50-50 chanoe. There is some evidence that the fault may be a 8 different kind, but there is also some evidence that i t wasn't 9 as long as it was to get it up to 7-1/2. So I wouldn't say 10 I'd clearly go in one direction on that. By that time they 11 may find the tau effect doesn't exist, but something else may 12 exist Well, the tau effect took this from.75 down to about 13 two-thirds, as ! recall, or six-tenths or something. But 14 there is something there that probably is going to show up, 15 anyway. 16 MR. LEWIS: Well, where are you, Chet? Are you 17 looking for guidance on what to do? I had 18 MR. SIESS: I am looking for a vote on 19 turned it back to the Chairman because ! am not a neutral in 20 this affair. I wanted the three people who disagreed with and it's now turned over to the Full Committee, unless 21 Dave 22 POSE would like to say something. 23 MR. BRAND: I don't believe so. 24 MR. AXTMANN: I've got a question. At the ( 25 subcommittee meeting someone pointed out that the Japanese go 1 - + - - - -, - -, -., -. -w.,- ,.-,e--, ,...--m,- --w- -,.,7, .,-,,.,,-,..y-.w,-,---,w--- ,,e---

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100 1 for stiff piping and Diablo Canyon calls for flexible piping. -s 2 MR. SIESS: No, nobody can have a simpler piping 3 system than Diablo Canyon. 4 MR. AXTMANN: Oh, excuse me. The statement is 5 the question was, why are we going from stiff to flexible? 6 MR. SIESS: There are moves aboard that will be 7 coming in to the committee on stiff vs. flexible piping, but 8 Diablo Canyon as it exists has got piping supported as stiffly 9 as anybody in the world, I would guess, 10 MR. AXTMANN: So we will be edified in 10 months? 11 MR SIESS: The proposed change is something we will 12 be debating. (" 13 MR. KERE: There is a two-volume report on a Staff 14 study of pipe stiffness and its relationship to seismic 15 behavior. 16 MR. SIESS: Yes. But this plant as built and as 17 rebuilt has got an awful lot of pipe supports. 18 MR. LEWIS: Chet, are you looking for an instruction 19 to go prepare a letter? Are you looking for a soft vote? 20 What do you want? 21 MR. SIESS: A soft vote would help. I can write a 22 letter with two paragraphs in it and we can debate it Saturday 23 if you wish. 24 MR. LEWIS: Well, we can certainly conduct a soft ( 4 25 vote if someone will make a motion on which we ought to vote.

i 101 1 MR. SIESS: Well, as chairman, I would move that you f~g 2 accept the majority opinion of the subcommittee and we write a 3 letter accepting the Staff's acceptance of the FRA at Level 1. 4 MR. LEWIS: Okay. The consequence of that, if 5 passed, is that you will prepare a letter which we will then 6 vote on. Okay. Is that a reasonable motion to vote on? We 7 have been talking about it now. Okay, let's have a show of 8 hands. Who would like to support Chet's motion? 9 CShow of hands.3 to MR. LEWIS: Opposed? 11 [Show of hands.3 12 MR. LEWIS: The majority, but three against. 13 MR. SIESS: Well, I will draft a letter and we will 14 see what it looks like. 15 Now I'm not sure that we can guarantee that the 16 Staff will accept our recommendation. 17 MR. LEWIS: Well, that seems to be a general rule of 18 life. 19 CLaughter.3 20 MR. SIESS: Well, it would be advice from the 21 Commission. 22 MR. MARX: I just wanted to find out if my 23 impression is correct. I have the impression that whenever an [V) 24 Applicant does a PRA, the Staff refuses to accept it without 25 doing it over again, and getting bigger risk numbers. Would

102 1 that be the case here? 2 ELaughter.3 3 MR. ISRAEL: Yes, that's part of the risk the 4 Licensee takes. 5 MR. MARK: Is the Staff going to be in 1988, when 6 this thing is with us, with enough money left to be able to do 7 a review of a Level 2 PRA? 8 MR. ISRAEL: Well, no, there is the one-week 9 evaluation, the two-month evaluation and the three-year 10 evaluation. It depends upon the money we have at that time. 11 MR. LEWIS: Okay. Are we done with the subject, 12 Chet? You have your instructions and you will write a report. Dave is missing 13 Okay. Gentlemen, the next item 14 for a few minutes, but he'll be back. 15 The next item on the agenda is recent events at 16 operating reactors. Do you want to head right into it or take 17 five minutes? 18 Five minutes. 19 CRecess.3 20 MR. LEW18: There are two options. We are coming up 21 to a group of, as I read the agenda, five or six five 22 recent significant operating events. I see two 23 possibilities. One, if we go through them in the order in 24 which they are listed, we will not get to Davis-Besse, which ( 25 many people think is the most significant one. The other

103 1 possibility is that we start with Davis-Besse and we may not Ol 2 get to the other four because it is one of the more 3 interesting ones, 4 The Staff is willing to do it either way, so we only 5 need a statement of preference from the Committee. What do 6 you want to do, start with Davis-Besse or end with it? 7 MR. REMICK: I support that. Start with it. 8 MR. LEWIS: Is that agreeable, to start with 9 Dasis-Besse? But at some point arbitrarily we will out it off to so that we get to the other ones. We will get only a 11 preliminary report on that now because the investigation team 12 hasn't come in. We will hear more about it next time. Okay. \\ 13 MR. JORDAN: My name is Rd Jordan. I'm the Division 14 Director Emergency Preparedness and Insident Response in ISE. 15 The five events that we plan to present to you are 16 selected out of eight by the Subcommittee on Operations, I'm sorry -- and these events we will go 17 selected out of 11 18 through as briefly as we can. 19 I would like to say just something about the 20 incident investigation program before we do regarding 21 Davis-Besse. 22 The Staff put together a Commission paper. This is 23 SECY 85-208, entitled " Incident Investigation Program," which ) 24 proposed to the Commission a manner of dealing with events, 25 and it is response to the Brookhaven study and also to the

104 l 1 ACRS findings. 2 The Commission has not adopted this; however, the 3 Staff advised the Commission in this paper that for the 4 interim the Staff was going to proceed with this method. The 5 date of this paper is June 10th. The Staff had prepared it, 6 and the event a Davis-Besse happened June 9th. So it was in 7 this fashion that we established this multi-discipline team 8 that is made up of technical experts from the"various NRC i 9 offices. 10 Ernie Rossi, who normally chairs this particular 11 meeting for I&E, is the team leader. He reports directly to 12 the Executive Director, Bill Diroks, in this fa2hion. A J 13 memorandum was prepared by Bill Dircks identifying Rossi as a 14 team leader and identifying J.T. Baird of HRR, Larry Bell of ,7 15 the 1&E Training Center, and Wayne Lanning of AEOD as the 16 experts in the various areas that seemed appropriate for this 17 particular investigation. 18 MR. LEWIS: I might just remind the Committee, for 19 people who worry about the ITT proposal, it is in your package 20 called 4.2, which is under Anticipated ACRS Activities. There 21 is a long proposal on that. 22 MR. JORDAN: Good. c 23 These team members were relieved from any other 24 assignments untti this particular task is completed, and that 25 tt scheduled to be done July 22nd with a presentation to the ~ t

l 105 i Commission immediately thereafter, and I would expect a 2 presentation to the ACRS as a specific event in your August 3 meeting. 4 The special investigative team is to prepare a 5 single report, and this represents the single investigation 6 that the Staff if preparing. That report would focus on fact 7 finding. It would identify the root causes and provide i 8 findings and conclusions, but it would not contain, normally, 9 recommendations. The recommendations subsequent to that would 10 he developed by the program offices responsible for the 11 various areas. If it impacted on inspection or impacted on j 12 licensing, then IAE or NRR, r e s p e c t i. " e l f, would be responsible 13 for coming up with those recommendations. \\ 14 The investigation was conducted or is being 4 15 conducted in a very factual manner. The very first effort was 16 to take statements from plant personnel who were involved on 17 shift at the time of the event, review strip chart records and 18 then go to procedures and logs and manuals, and finally 1 19 looking at the equipment. 20 They are in the process now of examining what they 21 feel may be the causes of the event, and it is obviously 1 22 multi-faceted. There was imposed a confirmation of action 23 letter that caused the plant insofar as possible to leave the 24 equipment in the "as-found" condition, which is one of the ( 25 issuesa that the ACRS, I believe, is interested in. l ., ~ - ..~...__.--....__.-..,,,__m_ -- -_~ m.._,

t 106 l l l h 1 The program is being administered by the AEOD l 2 office, but Dr. Rossi is reporting directly to the EDO in l l 3 terms of the findings. 4 So with that, I would request Al DeAgazio to give 5 the discussion and refer you then to the handout, and there is 6 page 12 of the handout that starts the presentation. 7 We did attach to the package an information notice l 8 issued July 8th to all licensees summarizing what we know l l l 9 about the event at this time. 10 A1. 11 Oh, yes, one other thing. We have received copies 12 of the licensee event report, and we will distribute extracts l 13 of that to the Committee. 14 Okay, A1. 15 MR. DeAGAZIO: I'm Al DeAgazio. I'm the project 16 manager for Davis-Besse in NRR. l 17 The sequence of events that I have has been put ( 18 together from information that has been available from the 19 team and information that was available frem the resident ( 20 inspectors. The sequence of events will leave out a lot of 21 the detail that is available but doesn't particularly affect 22 the sequence as we understand it now. 23 Before I get into the desoription of the event, I ) 24 think there are two packages that are being handed out to 25 you. One package has a slide which has the sequence of

107 /~N 1 events. That looks like this. ss 2 CElide] 3 That package has a diagram of the feedwater system, 4 but it's a little more simplified than the diagrams that 5 appear in the second package that is being handed out now. 6 The second package has the feedwater systems with the valves 7 in various positions, depending upon the condition of the 8 plant, and as we go through, I will refer to those. So if you 9 will, refer to the second package for the diagrams. 10 Before I go into the description of the sequence, 11 let me first describe what the feedwater system looks like at 12 Davis-Besse, or the feedwater systems look like at O \\ 13 Davis-Besse, and what the normal alignment of valves is. 14 The Davis-Besse plant has two steam generators, two 15 once-through steam generators. There are two steam-driven 16 main feedwater pumps. They can provide flow to either steam 17 generator. There are two turbine-driven auxiliary feedwater 18 pumps that provide auxiliary feedwater flow to either one or 19 both of the steam generators, depending upon valve alignment. 20 This plant also has a small capacity electric 21 motor-driven startup feedwater pump. This is used normally 22 just during sigt'up, and the later stages of plant shutdown. 23 The' capacity of the electric motor pump is less than one ( 24 turbine-driven auxiliary feedwater pump. The characteristio 25 curve of the pump is kind of flat, so it is difficult to say

108 's 1 whether it is half capacity or what it is because it is very ] 2 dependent upon the pressure in the steam generator. 3 So I think we will say it is no more than half of an 4 auxiliary feedwater pump as far as capacity is concerned, and 5 depending upon the pressure of the steam generator, it could 6 be considerably less. 7 MR. EBERSOLE: At the time of the occurrence of this 8 accident, was it not contemplated that additional pumps would 9 be installed at some point in the future? 10 MR. DeAGAZIO: There is a license condition that is 11 in effect now that would require them to provide a new startup 12 feedwater pump in a new location. The startup feedwater pump ~ ( i \\ 13 would have a capacity that is equal to one of the auxiliary 14 feedwater pumps. The piping arrangement for that startup 15 feedwater pump would not be the way it is shown here but it 16 would have the capability to also provide feed to the higher 17 level auxiliary feedwater nc zles in the steam generator. 18 MR. EBERSOLE: When was that condition imposed and 19 how long has it been more or less dormant? 20 MR. DeAGAZIO: We started working on that license 21 condition approximately last summer, about a year ago, and it 22 was finally put in place on the plant in January of this year, 23 after evaluation. The plant was in a refueling mode from -- I 24 think it was about September to January. 25 MR. EBERSOLE: But that was not operational when

_ _ _ _ _ ~ 109 1 this occurred? 2 MR. DeAGAZIO: The new startup feedwater pump is not 3 in the plant now yet. It is scheduled to be in, according to 4 license condition, before startup from the next refueling 5 outage, which would nominally be sometime next spring. 6 MR. EBERSOLE: Thank you. 7 MR. DeAGAZIO: During normal operation at power, the 8 feedwater pumps are providing flow through the feedwater 9 heating trains through a main flow control valve, and past the 10 isolation valves into the steam generators. The auxiliary 11 feedwater system, there are four valves that are closed so 12 that there is no open path from the pumps to the steam 13 generators. These are containment isolation valves. 1 didn't 14 These are steam emission valves from 15 have enough room and I w a's going to get complicated, so Point 16 A feeds here, Point B feeds here (indicating) The steam 17 admission valves are closed. Flow control for auxiliary 18 feedwater is by controlling the speed of the turbines. There 19 is not a flow control valve. 20 At the time just prior to the event, the plant was 21 operating at 90 percent power, and there were no special tests 22 going on at the time. The plant was stable. One main 23 feedwater pump was on automatic. One main feedwater pump was ( 24 in manual control. The purpose of that was that they had been 25 experiencing difficulties with the speed governors on the main

110 1 feedwater pumps. 2 At the last refueling outage, both of the feedwater 3 speed governors had been replaced with new models. The 4 initiating event was the tripping of the main feedwater pump 5 that was in automatic on overspeed. 6 The reduction in feedwater flow initiated a power 7 runback, and approximately 30 seconds later, the power level 8 had been down to 78 percent, but the runback wasn't fast 9 enough and a high pressure reactor trip occurred. 10 Just shortly after, about one second later, a steam 11 and feedwater rupture control system signal was generated, and 12 this signal was generated on low steam generator level, 13 apparently. It seemed to be spurious, and the cause of it is 14 speculated at the moment to be oscillations created by the 15 tripping of the turbine as far as the level control system for 16 the steam generators. 17 I'm sure that that's 'n o t at all firm at this point, 18 but that is the latest theory at the moment. 19 This trip cleared itself approximately three 20 seconds later. Whether or not it was related to that trip of 21 the steam and feedwater rupture control system, the main 22 steam isolation valves tripped. This valve and this valve 23 (indicating) () 24 This had the effect of stopping all steam flow from 25 the steam generators, and since the one main feedwater pump

111 1 that was running in manual at that point was deprived of 2 steam and yet it coasted down on the amount of steam that was 3 stored in the system beyond the isolation valves, it took 4 approximately four minutes for that pump to coast down and 5 trip. 6 MR. LEWIS: Could you remind me why the main steam 7 isolation valves have to trip at that sequence? 8 MR. DeAGAZIO: Why do the main steam isolation 9 valves trip in that sequence? 10 MR. LEWIS: It was normal that they should have at 11 that point in the event, or was it abnormal? 12 MR. DeAGAZIO: It really depends upon what kind of O 13 signal was generated from the steam and feedwater rupture 14 control system. First of all, there should not have been a 15 signal from the steam and feedwater rupture control system at 16 that time since steam generator levels were normal. But the 17 purpose of the steam and feedwater rupture control system is 18 to provide protection to the plant in the event of either a 19 steam line break, a feedwater line break. 20 It provides for starting of the auxiliary feedwater. 21 pumps in the event of low steam generator water level, and 22 provides actuation for the auxiliary feedwater system in the 23 event all four reactor coolant pumps are lost to promote N 24 natural circulation. 25 MR. KERR: But isn't the answer to Dr. Lewis'

112 ^^g 1 question that you don't know exactly why the main steam d 2 isolation valves closed at this point? 3 MR. DeAGAZIO: Exactly what kind of a trip was 4 generated, whether it was a half-channel trip or a 5 full-channel trip, I don't think we know at the moment. If it 6 was a full-channel trip, other things should have happened 7 besides the main steam isolation valves going closed. If it 8 was a full-channel trip, then some valves that should have 9 worked didn't work. If it was a half-channel trip, then the 10 main steam isolation valves should not have closed. 11 MR. LEWIS: Is that the answer to Dr. Kerr's 12 question? O 13 MR. DeAGAZIO: I think so. I think the answer is ss 14 we don't know just exactly why the main steam isolation valves 15 closed. 16 MR. EBERSOLE: Well, isn't it a fact, though, that 17 if you have a main steam line failure, it is obligatory that 18 you must shut one of these big boilers off and permit only one 19 to blow down on the pain of having excessive containment 20 pressure? 21 MR. DeAGAZIO: If you have a steam isolation break, 22 both of them are isolated. 23 MR. EBERSOLE: They are both isolated? () 24 MR. DeAGAZIO: Both of them. 25 MR. EBERSOLE: Both mains are isolated?

l l 4 113 l ~%g 1 MR. DeAGAZIO: Both mains are isolated. J 2 I am a little ahead of myself, but why don't I go ) 3 ahead and desoribe what happens on the steam and feedwater 4 rupture control system actuation on a low steam generator 4 5 MR. EBERSOLE: WEll, that is what gives you 6 redundancy to prevent discharge of both of them. You isolate 7 both of them. 4 8 MR. DeAGAZIO: That is correct. 9 [ Slide] 10 If there is a low steam pressure trip of the steam 11 and feedwater rupture control system, the main steam isolation 12 valves close on both steam generators and it doesn't matter e ~s 13 which steam generator has the low pressure as far as the main 14 steam isolation valves. 15 Oh, I have got the wrong diagram here. 16 [ Slide] 17 This diagram is for a low pressure trip on steam 18 generator No. 1. Both main steam isolation valves close. The 19 main feedwater trains are all isolated. The steam admission 20 valve for the feedwater pump that is normally dedicated to 21 f e e d 'i n g that steam generator is opened, so that steam 22 generator No. 1 -- I'm sorry. We have a break in steam 23 generator No. 1. Steam generator No. 2 is aligned to feed 24 to obtain its feed from aux feedwater pump No. 2. No. 1 is 25 not aligned to No. 1 steam generator.

114 1 The steam admission valves from that steam 7-'s %,) i 2 generator, which presumably has a break, have not opened. The 3 steam admission valves from the No. 2 steam generator are 4 opened. That is this valve (indicating). And the discharge 5 flow from the No. 1 pump is aligned to feed the No. 2 steam 6 generator, as is the No. 12 feedwater pump. 7 So that this steam generator, if it has the break, 8 if it is the one that has the low steam pressure, it is 9 totally bottled up and feedwater is provided only to No. 2 i 10 steam generator. If No. 2 fails, has the low pressure, then 11 the No. 1 steam generator gets all the feedwater flow. 12 If beth of them see a low pressure signal, they will / i (m,/ 13 both get isolated. There will be no steam admission, no 14 feedwater flow path until one of the steam generators 15 repressurizes. At that point the signal clears and the one 16 that recovers pressure will get the feed. 17 [ Slide.] 18 Since we are discussing the flow paths on a low 19 steam generator water level, or on a high feedwater to steam 20 pressure differential which would be indicative of a feedwater 21 line break, the action is similar but a little different. 22 The isolation valves to the steam generators are not 23 closed, so that feedwater flow can be provided from the pump 24 that is dedicated for that particular steam generator. Steam ) 25 admission to the auxiliary feedwater pump turbines is also

115 x 1 from the respective steam generator. .2 Going back to the sequence now, the main steam 3 isolation valves had closed, all main feedwater flow is off, 4 at this point the steam generator water level is declining. 5 I'm sorry, I'm getting ahead of myself. 6 We had the spurious trip of the steam and feedwater let me 7 rupture control system, and that had the effect of 8 back up. I am still in the wrong sequence. 9 The spurious trip had the effect of closing the main 10 steam isolation valves and cleared approximately four seconds 11. later. The main feedwater pump took approximately four 12 minutes to coast down and at that point the steam generators 13 are without feedwater. 14 At about six minutes into the event, there was an 15 actual, a real low level steam and feedwater actuation signal 16 generated for steam generator No. 1, and that aligned the 17 valves as shown here [ indicating 3, forgetting the fact that it 18 was No. 1. 19 I'm sorry, it didn't matter. 20 The valves were aligned as they are shown here. The 21 operator, recognizing that he was getting into a situation 22 where level was dropping, attempted to initiate auxiliary 23 feedwater and not depend upon the steam and feedwater rupture (J I 24 control system, and the automatic system beat him by just a 25 second or so, but when he made his action, he unfortunately

116 1 i selected the wrong buttons on the control panel and selected 2 the low pressure trip instead of the low water level trip and 3 so it had the effect of aligning the valves this way. 4 [ Slide.] 5 But since he punched the buttons for both steam 6 generators, both steam generators were isolated from feedwater 7 and there was no feedwater flow to either steam generator. 8 These valves were closed also, and these and this valve was 9 closed along with this and this [ indicating]. 10 So there was no flow path into the steam generator 11 and the steam generator water levels are now going down. At 12 this point they are drying out and approximately one minute \\ws 13 later he recognized that he made the error and tried to 14 correct it. 15 When he did correct it, what should have happened 16 was that the valves that were closed should have opened 17 according to this diagram. 18 However -- 19 CSlide.] the system should have aligned this way. 20 21 However, this valve and this valve failed to open. 22 MR. LEWIS: Could I just jump ahead and be sure that 23 by the time you are finished, you will tell us how many ) 24 equipment malfunctions there are or there were because I'm 25 having a little trouble distinguishing what is normal response

117 1 to the event and what is abnormal. 2 MR. DE AGAZIO: This is a very complicated system, 3 and I do have a list of what we know has failed so far. 4 The operators then were dispatched throughout the 5 plant to attempt to get these pumps started again, which 6 tripped just shortly after they started on the first real low 7 level signal from the steam and feedwater rupture control 8 system to attempt to open these valves and to attempt to get 9 the electric motor-driven start-up feedwater pump in 10 operation. And this was a manual operation. You have to go 11 to the valves locally and manually open the valves. 12 Also, to replace some control fuses which were b 13 pulled to prevent starting of this pump against a closed 14 suction valve. By approximately -- 15 MR. OKRENT: Excuse me. Were the positions of those 16 valves on the feedwater line indicated in the control room? 17 The two that opened. 18 MR. DE AGAZIO: These valve positions are all 19 indicated in the control room. 20 MR. OKRENT: Thank you. 21 MR. DE AGAZIO: This one is not [ indicating 3. The 22 others are just manual local valves. 23 The operators were successfull in getting the D 24 start-up feedwater pump started about the same time they l 25 were successful in resetting the turbine-driven steam water -- i

118 1 turbine-driven feedwater pumps, so that by about 16 minutes or 2 so they had started restoring feedwater to the steam 3 generator. 4 J u s f. prior to restoring feedwater to the steam 5 generators, the pressure in the reactor coolant system had 6 risen to the point where the PORVs had actuated. They 7 actuated three times. There was indication that the FORV 8 failed to recede on the third time. Whether this was actually 9 a failure of the valve or not, I think there might be some 10 question, because just about that time the feedwater was being 11 restored to the steam generators, so whether the drop in 12 pressure was what caused the operator to believe that he had a 13 stuck-oven PORV or whether it was some other indication, we l 14 don't really have that information at the moment. 15 The latest information from the site is that the 16 valve has been disassembled and there was nothing abnormal 17 about the valve. So the reason why it stuck, if did stick, is 18 not known. 19 Once the feedwater had been restored, the plant 20 entered a normal cooldown. 21 [ Slide.] 22 There was at seast 13 or 14 different malfunctions 23 or failures or unexpected occurrences. 24 First of all, the initiating event was the main 25 feedwater trip on overspeed.

119 -s 1 Following that was the unexplained either spurious 4 2 half trip or full trip of the steam and feedwater control 3 system. The unexplained action of the two main steam 4 isolation valves which should have closed, if this was a full 5 trip, should not have closed if it were a half trip. 6 Not shown on here is if this were a full trip of the 7 steam and feedwater control system, then other valves should 8 have actuated which did not. 9 The two auxiliary feedwater pumps operated for a few 10 seconds after the initial low level signal from the steam and 11 feedwater rupture control system. Then they tripped out on 12 overspeed. 13 Incidentally, one of those speed governors on the 14 auxiliary feedwater pumps had been replaced at the last 15 refueling outage with a new design, and the other one was the 16 original design, but they both did trip out. 17 The two isolation valves in the auxiliary feedwater 18 line that failed to open when reset -- 19 MR. LEWIS: Is it known that they got the signal to 20 open? 21 MR. DE AGAZIO: Yes, when the operator recognized he 22 had made an error and corrected his mistake, he pushed the 23 reset button which should have removed that signal and caused ( 24 the valves to open again. 25 MR. LEWIS: But it isn't known whether the valves 4

l 120 i got the signal; it is only known that he pushed the button? I f-- N,g I l 2 am trying to find out whether it was a valve failure or a 3 control failure. 4 MR. DE AGAZIO: I think the indication was that it 5 was a valve failure, not a control failure, because when the 6 operator went down and moved the valve to slightly off the 7 seat manually with the hand wheel, then the valve continued to 8 open. 9 The FORV failing to reseat -- and there is a there was 10 question mark that might be associated with that 11 a start-up feedwater valve that failed to open and had to be 12 opened manually. After the auxiliary feedwater pumps were i 13 restarted, one speed governor failed to respond to the control 14 from the control room and had to be operated manually. 15 When the auxiliary feedwater system was restored, 16 there was a low suction on the system and it caused a 17 switchover to a back-up service water supply, and somewhere in 18 the sequence was a damaged turbine bypass valve. l 19 MR. LEWIS: You know, I am an ignorant person, but 3 and of course the final report on this isn't out yet, 20 if t 21 but if I really am to believe that on a challenge produced by 22 the tripping of one main feedwater pump, there were t 23 subsequently nine independent failures of components in the l 24 following sequence, I would take that very seriously. You 25 know, I reserve judgment until I know what actually happened,

121 i but that would have serious implications. -s J 2 MR. EBERSOLE: Yes. In that connection, do you 3 intend to put together an analytical study of what is the 4 probability of this combination of events and use it as a 5 backdrop to consider PRAs in the future? i I 6 MR. DE AGAZIO: I'm not sure a decision has been 7 made on that yet. I think that until the report is issued by 8 the team and we know more about this with certainty as to what 9 happened and what failed, I think it would be premature. 10 MR. EBERSOLE: It may be that this is a milestone 11 occurrence in this context, and that we have to look at PRAs 12 in a somewhat less optimistic manner than we currently do. ( 13 MR. LEWIS: Well, Jesse, just in defense of PRAs, we 14 should wait until we find out what happened. There is sero 15 probability that there would be 10 independent failures. So 16 we will find that out. ~ 17 MR. MARK: Maybe this is a precursor for TMI-1 18 MR. LEWIS: It may well be. In fact, since this is 19 the first time the IIT system has been exercised, even before 20 it's been approved by the Commission, it would be very 21 interesting to see if they can get at the root cause of this 22 event. 23 MR. JORDAN: Yes, and that is the purpose, of () 24 course, of the way the team is constituted and managed. I 25 sincerely hope that works, i 1

122 l "'g i MR. LEWIS: I am holding my breath. 2 MR. REMICK: The title there says known equipment ) 3 failures or malfunctions, and you have the main feedwater 4 trip. Was that a malfunction? You said it was due to 5 overspeed of the turbine that was on automatic. Was that a i 6 spurious malfunction, or was there a reason for the overspeed 7 trip? 8 MR. DE AGAZIO: The plant had been experiencing 9 difficulties with both of these main feedwater pumps the prior 10 weeks, so I count that as a malfunction in the plant. It is 11 the initiating event here, but they had been having 12 considerable difficulty with these pumps, which is the reason 13 why they were running one of them at manual, so they could 14 minimize 15 MR. KERR: It's my impression from another source, 16 and perhaps mistaken, that they had indeed called new control 17 systems on these pumps recently. Is that not the case? 18 MR. DE AGAZIO: I'm sorry. 19 MR. KERR: I got the impression from another source 20 that they had recently installed new control systems for both 21 feedwater pumps. 22 MR. DE AGAZIO: Both main feedwater pumps had been i 23 equipped with new speed governors on the last.eutage. i 24 MR. KERR: And they were having some difficulty with 25 getting them adjusted or making them work properly?

n l i 123 1 MR. DE AGAZIO: They were supposed to be improved f 2 models, and one auxiliary feedwater pump had also been I l 3 equipped with a new speed governor. 4 MR. REMICK: Do you know if the pump actually did 5 overspeed and that is the malfunction and the trip worked, or 6 the pump didn't overspeed but tripped by a falso signal that 7 was overspeed? Do you know which? 8 In other words, was there an actual overspeed of the 9 turbine pump, of the turbine and pump? 10 MR. DE AGAZIO: I don't believe I have that 11 information. 12 MR. LEWIS: Forrest, there was a comment in the LER \\' 13 which has a lot more information in it, which I can't find at 14 the moment, and I don't know whether it's the answer to your 15 question, that somewhere there was frequency to voltage 16 converter trip that had failed on an overspeed control car. 17 And I may not be quoting it exactly, but there was something 18 like that. 19 MR. EBERSOLE: How many minutes were left before 20 core damage would have occurred? 21 MR. DE AGAZIO: I'm afraid I don't have the answer. 22 MR. SHEWMON: There was almost no loss in the 23 primary system, was there? ) 24 MR. DE AGAZIO: Not much, a blowing down three times 25 to the FORVs for a short period of time. So at that time they

m i 124 1 hadn't lost ruch, anyway. 1 i -/ \\ 2 MR. SHERON: Brian Sheron from the Staff. 3 This leads obviously to the question about ~ 4 feed-and-bleed on this plant, what mitigative actions could 5 have been taken. We have done a fair amount of hand 6 calculations since the event and we are also in the process of 7 doing more sophisticated calculations with RELAP-5 right now, 8 and we are going to report this to the ECCS subcommittee I 9 think on the 31st. 10 But what we determined is that if the operator had 11 taken no action whatsoever to either turn on makeup pumps or 12 turn on the start-up feed pumps, open a PORV, we probably i \\ 13 would have seen core uncovery in about 50 minutes, five oh, 50 14 minutes. Generators dry out very shortly in a few minutes, first it is displacement of 15 and the rest of it is just 16 water by steam collecting in the high points, until you 17 uncover the vent path, which is the surge line to the 18 pressurizer. And then it is just a boil-off. 19 We have also looked at the capability -- we also 20 looked at the capability of what the makeup pumps, the 21 start-up feed pump and the FORV could have done, and we have 22 generated a table that's what you find out is there is a 23 matrix of a combination of things that could have been done. I'm sorry, both 24 Had they started both feed pumps 25 makeup pumps, opened the PORV and turned on the start-up

125 (^s 1 feedwater pump, they most likely would have been able to keep 2 the core covered. There was sufficient capacity there to 3 remove decay heat. l l 4 The calculations also indicate that if they had just 5 turned on the two makeup pumps and opened the PORV and had not 6 even started.the start-up feedwater pump, that they still 7 could have kept the core covered. 8 If you just had one make-up pump, you would probably 9 have uncovered the core, but you would have extended the time 10 considerably. 11 Also, remember the event occurred at 90 percent 12 power, so a lot of these numbers also will change at 100 13 percent power. 14 In terms of how much time did an operator have to I 15 initiate these actions, right now what we see is that and 16 these are just various numbers we have -- if the plant had i 17 been operating at 100 power, rather than 90, if they had l l 18 initiated just two makeup pumps and no start-up feed pump, if l l l 19 they had initiated makeups at the time of steam generator i 20 dry-up, which I think is about two or three minutes, they 21 wouldn't have uncovered the core, but had they waited 20 22 minutes to initiate the make-up pumps, they would have gotten 23 core uncovery. 24 I beg your pardon? 25 MR. ETHERINGTON: With damage to the core or not?

) 126 ) 1 MR. SHERON: Well, what we are calculating is core j p-- \\w,) i r 2 uncovery. Typically damage occurs some time 15 or 20 minutes 1 3 later. Once you uncover say about half the core and you start 4 getting substantial clad heat-up -- we haven't done the I 5 calculations into the core. These are just until when the I 6 core uncovers. 7 MR. REED: Your numbers, most of them surprise me. 8 Your 50 minutes surprises me. Your two to three minutes to 9 dry out on the secondary side of the steam generators does not 10 surprise me, because that number has been floating around a 11

lot, 12 And, of course, I was quite surprised that the 13 system went for 11 minute without secondary feedwater, and 14 apparently there was no core damage and you didn't get above 15 594 degrees.

16 Now I have to conclude that the reason here that you 17 are giving me such numbers as 50 minutes to top of core is 18 because this particular B&W plant has high set steam 19 generators vs. the others with low set steam generators, and 20 therefore you are getting the drain-down advantage of the loop 21 piping in the steam generator primary side; is that correct? 22 MR. SHERON: No. I asked the very same questions of 23 the individual who did the calculations, and just to lend a ) 24 little bit of credibility, the person who did the calculations 25 worked for about 10 years for B&W in the analysis area, so

127 ] 1 he's D 2 MR. REED: I think you already answered my question. 3 MR. SHERON: The answer is that volume of the core, 4 as I understand it, comes out about the same. So the raised 5 loop plant really doesn't behave that much differently than 6 the lowered loop in terms of time to core uncovery. 7 My understanding is, it also has to do with the 8 relative location of the surge line on the hot leg. During 9 the boiloff process, once you've dried out the generators and 10 you saturate the primary system and you generate steam in a 11 primary system, that steam will first collect in the top of 12 the vessel, and what it does is, it displaces water. And the \\s/ 13 displacement process is very quick, and so you're pushing 14 water right out of the pressuriser. 15 Once you fill the upper head, you then collect the 16 steam at the top of the candycane, and you still displace 17 water until you uncover the surge line. Once you uncover the 18 surge line, then steam can escape out the pressurizer, and 19 then it becomes just a boiloff process. 20 MR. REED: All right. What I think I'm seeing -- 21 and every time I see this PORV thing, I see one PORV; I don't I see one series, in-series block 22 see plural here, PORVs i 23 valve; is that correct? 24 MR. SHERON: Yes. 25 MR. REED: That is pretty thin ice for getting an

128 1 exit path, isn't it? I mean, he closed the PORV. Suppose he (S N-2 never could have gotten the block valve, suppose he never got 3 it open again? 4 MR. SHERON: That's correct. Again, you have to 5 remember that the whole issue of feed and bleed and the fact 6 that B&W plants -- and this plant is not unique in that all B&W plants only have one PORV. You know, the 7 respect 8 PORV, HPI, or makeup system was not originally designed for 9 feed and bleed, and the PORVs were put.on these plants, as you 10 well know, just to protect the lifting of safety valves. 11 So you're correct in the sense that obviously a 12 plant that has a higher PORV capability with more PORVs does l (* 1 y 13 have, perhaps, a much higher capability of feed and bleed. ( x 14 MR. MARK: I got the impression from reading one of 15 these things that there was pretty hectic fifteen minutes in 16 that control room, and with the exception of the one operator 17 error, it looks as if they must have done everything just 18 about right and very quickly. Is that the proper impression? 19 MR. DeAGAZIO: That's reasonable. I think it was 20 not only hectic in the control room, but to recognize that to 21 start up the startup feedwater pump, the auxiliary feedwater 22 pump and get the valves open to both the startup feedwater 23 pump and the stuck-open valve or the stuck-closed valves in ( 24 the auxiliary feedwater system path took a lot of running 25 around in this plant. They had to go to vital areas with key

129 -w 1 access, card access. h 2 MR. MARK: They did a number of things within a 3 period of two or three minutes. 4 MR. DeAGAZIO: In about nine to fourteen minutes, ) 5 they were moving through that plant quickly. 6 MR. MARK: Here is a list of things which 7 malfunctioned, and there was one operator error. Sometime it 8 might be worth saying that this crew was rather good and did 9 very well. 10 MR. DeAGAZIO: With regard to the operator error, 11 the control panel where he made the error looks like this, and 12 I could pass this around. m 13 MR. MARK: Yes That was at fault. 14 MR DeAGAZIO: This has been identified in the 15 control room design review as a human engineering defect. 16 What he did was, he punched the two top buttons, when he 17 should have hit.this one (indicating) 18 MR. KERR: I thought the aux feedwater system was 19 automatic start; is that not the case? 20 MR. DeAGAZIO: The auxiliary feedwater system is 21 automatic started by the steam and feedwater rupture control 22 system. 23 MR. KERR. Well, why did he have to punch anything? 24 MR. DeAGAZIC: It's my understanding -- 25 MR. WARD: Isn't that just confirmatory?

130 1 MR. DeAGAZIO: It's really not to depend upon the 'u 2 automatic signal, but tc, when you recognize that you're going 3 to need an action, to taxe the manual action-and not depend 4 upon the automatic action. 5 MR. WARD: It is just sort of follow-up, which is, I 6 think, in most plant procedures, and unfortunately he followed 7 up with the wrong button. 8 MR. DeAGAZIO: I think that what happened is that if 9 you know where the location of this panel is with respect to 10 the indicators that he was looking at, he had to move around 11 the control room, around from the central horseshoe and get 12 around to the back panel, and the automatic signal just caught l 13 it first. 14 MR. KERR: I'm not trying to be critical. I'm 15 trying to understand. 16 My impression was that after TMI, the Staff decided 17 that aux feed systems should have automatic startup, because 18 this was likely to produce less operator error than manual 19 startup. 20 Now what you have here is a combination of the two, 21 it seems to me. You still have the possibility of operator 22 error, even though you have automatic startup. Maybe there 23 should be a rule that says once you need aux feedwater, you I \\ 24 should handcuff the operator for ten minutes, so that he can't 25 foul things up, because he did, I think, what any normal

131 i 1 operator or well-trained operator would do. He tried to do lO 1 2 something, even the wrong thing. 3 MR. LEWIS: But you know, they are telling us, Bill, 4 that the operators made one mistake, but the machinery made 5 ten. 6 MR. KERR: But we don't know whether the machines 7 made mistakes or not. if I have to predict 8 MR. LEWIS: Well, I really 9 anything, it is that the final story on this will be quite 10 different from this preliminary version that we are hearing 11 now. J 12 MR. EBERSOLE: In the Palo Verde case, they are 13 counting on secondary blowdown. 14 MR. SHEWMON: Wait. Let me stay with that for a 15 minute. The operators' procedures are, " Don't see if the 16 machine is working; punch' buttons," is that it? 17 MR. WARD: As I understand it. I don't know, 18 Glenn. Almost all emergency procedures call for confirmatory 19 actions to follow up presumed automatic actions. And I gather 20 in a lot of cases, instead of looking for an indication that a 21 valve is open or a pump is started, the confirmatory action 3 22 consists of doing manually what's necessary to open the valve 23 or start the pump, 24 Now, you know, I'm not arguing whether that's the s 25 right thing to do or not. I _,~

J'- ( fx } 132 / 1 A 1 MR. KERR! ft may. be the right thing. I was just (/I i, .c x. 2 curious. 3 _MR. WARD: Well,gthat's what's pretty st$ndard. 4 MR. LEWIS: I'wi 1 confirm that. l' 5 6 MR. WARD: Well, this is a case where it looks like 1 / 6 that is not a good way to do it. s 5 J v --*U 7 MV. LEWIS-I wi11 confirm that. I have checked some '( [ e 8 specific o r. e s recently, and,the procedure says, " Verify," and e s t 9 what the operate.rs are tiught is that verify" means push it 10 manually to make sure it happened. J.. 11 MR. KERR: 'Well,, ram the long run, that maf be the 'I t t. f...) < j.'. p'2 thing to do. p is sort of like with a scram. You 4~ [It 13 MR. WARD t } 14 know, there irfsome argument and discursion after S a'J a m, 1 l e 3 p f i whenfthere's an automatic s c r.s m, should the operator 15 whether s ' h A 4 b u't t'o n ? ') 16 instantly h f. t the manual scram 17 MR. SHEWMOPr: In that case, it was the right thing. L-18 M R'. ' W A R D : Yes. There, it would have been right, s 2. ~h i 4 19 MR. REED: 7 would like to support 4 little bit what 20 Carson Mark said. ,I t looks like the oper.atordidid a good job "y 21 here, except for the one error, which they are entitled to, s ,a is the basis of design from way back, equipmentifailure 22 [which f j-humar iailure. 23 and one V 24 I look at this design, and I've been shocked ever I can't ih.agtne why ^ 25 since I'1ooked at it, because I don't t;

133 1 steam-driven pumps were used as auxiliary boiler feedpumps 2 here'and used only. That just bothers the hell out of me as a of course, this is a 3 design i ssue, and I don't understand 4 very complex valving arrangement, automated and otherwise, 5 with respect to introducing water into one or the other steam 6 generators. 7 And the other thing that bothers me is, I see all 8 kinds of closed valves. And as Jesse knows, I mentioned that 9 with respect to System 80. Valves have habits of not 10 functioning, and they are quite unreliable. And whenever you 11 can create systems with checkvalves, without closed valves, 12 your much better off to do that from a reliability of 13 performance point of view. 14 So I see a lot of design things that bother me, and 15 I shall be listening to the final report very closely to see 16 whether the design issues-are really raised this time around. 17 MR. EBERSOLE: Well, let me enlarge a little bit on 18 that, Glenn 19 These turbine-driven pumps unfortunately are also 20 buried i n the bowels of the plant and are cooled by AC-driven 21 environmental cooling fans, which makes then interdependent 22 with the AC system, unless that's been fixed, and I don't 23 think i t has, i 24 MR. DeAGAZIO: I verified that. It is AC only. 25 MR. EBERSOLE: So either electric power failure or m - --u_..,.,,

134 g i steam failure or any other kind of failure can knock these l \\_Y 2 pumps down. It is a design disaster. 3 MR. WARD: In addition to wh'at Carson said, isn't it 1 4 fair to say that after the initial error the operators made, I 5 they seemed to fully understand what was going on? I mean, in 6 some other incidents we have had, they have been compounded i 7 because the operators in the control room didn't understand l l 8 what the condition of the plant was. l 9 Is it fair to say that here they understood it all 1 10 the way through? i 11 MR. DeAGAZIO: I haven't heard of any information 12 that would indicate that they did not understand what was 13 happening and that their actions, at least as far as the 14 sequence is concerned, indicates that they did understand. 15 MR. SHERON: Excuse me. I have no idea what you're 16 saying. 17 MR. DeAGAZIO: I was saying, there is no information 18 that would lead me to 2.41, eve that they did not understand 19 what was happening t w s ue r the question specifically. It 20 is my understanding that they were, at that point, following I. 21 that the ADOG procedures were in place, and they were i 22 following those procedures. 23 MR. EBERSOLE: It's fair to say, however. that this ( 24 plant meets all regulatory requirements, doesn't it? 25 MR. DeAGAZIO: I believe so, yes.

135 1 MR. JORDAN: With respect to the operators and their s \\ 2 understanding in the control room, we were disappointed in the 3 Operations Center in Bethesda with the information we received j 4 from the plant. We were called 36 minutes after the turbine 5 trip, and we were advised that the plant was stable. They had i 6 had a main feedwater. pump trip. We were unable to get details 7 for-some time, and it was a third phone call before an unusual 8 event was declared, which would be the lowest of the emergency 9 classes. And the Licensee did identify in this third phone 10 call with the Shift Supervisor that they were, indeed, in a 11 site area condition, although they had not declared it during 12 that timeframe. I i 13 So the individual we were talking to was the Shift 14 Technical Advisor. The Shift Technical Advisor at this plant 15 is on a 24-hour call situation in a building outside of the P 16 control room area. He came into the control room area about 17 ten minutes into the event and really never got caught up in 18 terms of providing the Operations Center with anything like 19 the obvious concerns that the personnel that were actually 20 coping with the event had. 21 So we don't -- 22 MR. KERR: But if you had to choose between their 2 23 looking after the accident or informing the Information 24 Center, you would sort of take the former, wouldn't you? g j 25 MR. JORDAN: I'm sorry, but I have to insist on t 1

136 ["'g i both, that if this event had gone awry, then the NRC would not 5 2 have been in a position to initiate the emergency response 3 capability and make determinations as far as what offsite 4 measures should be taken. 5 So we do really need both, and it wasn't provided in 6 this case. 7 MR. LEWIS: Wait a minute. Aren't emergency 8 measures de t ermined -locally ? They didn't declare an 9 emergency, and that is bad, but doesn't that declaration 10 trigger the emergency procedures locally? 11 MR. JORDAN: Yes, it does. i 12 MR. LEWIS: Okay. It doesn't have to go through \\ 13 Washington. 14 MR. JORDAN: No. The Licensee identifies the extent 15 of the emergency. He classifies it. The information to the 16 NRC gives an opportunity for a second look at whether he has, 17 indeed, classified it correctly and whether the appropriate 18 actions are being initiated. 19 MR. LEWIS: Yes, but that's a surveillance role. 20 You used the word " initiation." 21 MR. JORDAN: I'm sorry. In this case, it was okay. } 22 But I'm just postulating that if it had gone sour worse than 23 it did, if they had not gotten the feedvater back on, then the b 24 agency was quite behind in its response. q j 25 MR. REED: How many minutes did you say it was J

137 1 before you got an understanding of the event off the red {} %./ 2 phone? How many minutes was it? 3 MR. JORDAN: The initial call came to us 36 minutes 4 after the trip. Our actual reasonable understanding didn't 5 occur until later when the Resident Inspector got to the i 6 site. We had only bits and snatches of what had happened. So 7 it was an hour and a half or so after the event. ? 8 MR. REED: Well, did they meet the regulatory 9 requirements? 10 MR. JORDAN: Well, I mean in terms of that the plant 11 was stable, that the knew what had transpired from the 12 superficial respect. 13 No, we' don't know what happened yet. That's 14 correct. f 15 MR. REED: Did they meet the regulatory requirements 16 as soon as possible? 17 MR. JORDAN: They met the regulatory requirements. 18 MR. REED: Within one hour, right? 19 MR. JORDAN: That's correct. I'm only indicating 20 that from the agency's knowledge, what was conveyed to the NRC 21 in 36 minutes was not representative of what the operators-( 22 themselves knew, clearly, and that's all that I'm trying to 23 get across. 24 MR. REED: I'm still going to vote with the 25 operators, that they should have concentrated absolutely, as 4 1 j

138 } 1 they did, and let the red phone come second, as long as they v 2 met the regulatory requirements. 3 If you're on an airplane going down, you are not 4 going to necessarily get on the red phone. You would rather 5 tend to your airplane going down. 6 MR. JORDAN: I agree. But if it crashes, we'd like 7 to know about it. It's too late; that's right. 8 MR. REED: Well, we hit the ground. You can recover 9 the data boxes weeks later. 10 [ Laughter.3 11 MR. WARD: Okay. Thank you very much. 12 Jesse, we are going to lose five of our members at b) s 13 6:30, so I think we really ought to plan on ending today's 14 meeting at that time. We have about -- Mr. Fraley, will you 15 need a full half-hour? 16 MR FRALEY: No. Fifteen minutes ought to be 17 plenty. 18 MR. WARD: So do you think we could point toward 19 finishing this item at about 6:15? 20 MR. EBERSOLE: Well, all I can ask is for the 21 individual presenters to expedite it, I guess. 22 MR. JORDAN: We will go ahead in the order listed. 23 MR. EBERSOLE: I think you can march quite quickly _,/ 24 through this. This is an absolutely unbelievable system 25 interaction.

139 1 MR. JORDAN: This is the Hatch-1 stuck-open relief 2 valve. George Rivenbark'of NRR will give the presentation. 3 He is the Licensing Project Manager for Hatch. 4 [ Slide.3 5 MR. RIVENBARK: Well, this was on May the 15th and 6 it happened in the evening. Unit 1 was operating at full 7 power. It is surmised that a crane was passing overhead and 8 ruptured a line that provides water pressure to the charcoal 9 filter deluge valves, the fire protection valves. When the 10 crane hook drug across a line, it cracked it, bled pressure 11 off of the line. 12 This pressure normally kept the valve closed that 13 caused the valve to open that spray water into the charcoal 14 filters. The operators did not know this at the moment. They 15 knew it some minutes later, maybe 15 or 20, possibly, when 16 they saw water dripping into the control room through the air 17 conditioning ducts. 18 MR. EBERSOLE: This crane was out in the turbine 19 hall, was it not? 20 MR. RIVENBARK: The crane at the time was passing 21 from one turbine building to another turbine building, and it 22 goes over the control room. And above the control room, on 23 the floor immediately above the control room is the location ( 24 of the air conditioning and the equipment. and this has no protection against outside 25 So

r-140 1 influences. It is open to the crane area, and so the crane 2 hook which was presumed to have been drug across there, bumped 3 this line and cracked it. 4 As a result of the water dripping into a instrument 5 panel, it happened to be into one of the newly installed 6 analog transmitter trip system panels, and it resulted in one 7 of the SRVs opening. The SRV opened several times and closed 8 and then opened and stayed open. 9 At the point that the SRV opened and stayed open, 10 the operators scrammed the reactor. After the reactor was 11 scrammed, the feedwater pumps -- initially when it was 12 scrammed because the SRV was opened, the level went down, the 13 feedwater pumps quickly recovered that level, 14 They tried to close the SRVS by pulling fuses, but 15 the procedures were incorrect, they had the fuses listed 16 incorrectly, and before they could finally locate the right 17 fuse some 30 minutes into the event, the SRV closed by itself, 18 and they don't know why. They don't really know exactly why 19 the SRV opened. 20 They attribute

it naturally to the moisture getting.

21 into the instrument panel The moisture that went into the 22 instrument panel, in addition to just causing the SRV to open, 23 also caused eventually one of the power supplies that were in 24 there to burn out. 25 The reason that the water dripped into the control

141 1 room was determined -- well, aside from the fact that the 2 deluge valve went off, was that the vents that are in the and the plenum looks like roughly this. 3 plenum 4 [ Slide.] 5 The air conditioning plenum. The charcoal filter -- 6 this is in the room above the control room, immediately above 7 it, and here the ducting comes in from the outside of the 8 building air intake, and here is the charcoal filter sitting 9 here, and then the discharge line out into the control room 10 goes this way and the intake back from the control room comes 11 immediately into the edge here. 12 This is pretty much a straight drop into the control Ot) N 13 room. 14 Well, these drains which were located at the bottom 15 of this box were plugged and the water filled up to this 16 point, and at this point it flowed over this way and down into 17 the control room [ indicating 3. 18 The action that was taken subsequently was to dry 19 out the instrument panels and replace the power supply and 20 clean the air conditioning -- the drains for this particular 21 box and they checked the auxiliary box and it did not have its 22 drains plugged. 23 So ( 24 MR. EBERSOLE: It would appear that just without the l' 25 crane dragging its parts around, that there is a potential for l

142 rN 1 liquid egress from a system inside there, and therefore a ( 2 straight potential to run right down into the control room 3 from a water pipe, which is almost close to the old classical 4 scenario of having a toilet on the ceiling which overflows 5 after you flush it, and the whole set-up seems to be loaded. 6 Is that a characteristic of the general condition at 7 Hatch? 8 MR. RIVENBARE: If the drains plug up and the water 9 floods into the box, it is going to run over. 10 [ Laughter.3 11 MR. LEWIS: The safety relief valve is electrically 12 operated from that panel, so when you say you don't know what T 13 caused it, you mean you don't know exactly 14 MR. RIVENBARK: They don't know how the water 15 affected the circuits, what it did or exactly how it caused it 16 17 MR. LEWIS: But the control for the circuit is at 18 that point? 19 MR. RIVENBARK: Oh, yes. 20 MR. EBERSOLE: Don't we need to get the water 21 ingress potential completely away from the panel board? Not 22 just patch up the drain holes and all, but just get the 23 potential out. And aren't there easy ways to do that? 24 MR. RIVENBARK: I'm not prepared to answer that. 25 MR. EBERSOLE: I guess what is going to be done is

143 1 just what you said, you leave it cocked again for this. 2 MR. RIVENBARK: Would you repeat that, please? 3 MR. EBERSOLE: I'm saying what I heard you say, you 4 are going to fix the drains that were plugged 5 MR. RIVENBARK: No, you didn't hear me say I was 6 going to fix it. What you heard me say was that the drains 7 were cleared, and having fixed the cause and put everything 8 back into working order, the plant was allowed to restart. 9 MR. EBERSOLE: Yeah, right, so you fixed the drains. 10 MR. RIVENBARK: I do know that they are looking at the plant is looking at the possibility of 11 the possibility 12 removing the water altogether from the plant, 13 MR. EBERSOLE: Good. That's all I was after. 14 MR. REED: I want to make a point here, because we 15 frequently talk about valves and malfunctioning. I believe he 16 said that the safety relief valve, even though it got the 17 signal, the fuses had been pulled or something, to go closed, 18 it didn't. 19 I want to make a point that this was in -- even 20 though this is a BWR, without boron in the water, that this 21 was an internal pilot operated electromagnetic relief valve, 22 and I have great concerns about their reliability on PWRs, but 23 I find on BWRs they can be reliable. [ 24 MR. RIVENBARK: I would like to correct one little ( 25 point. I did not say that they pulled the correct fuse. As a _ _ _ ~ _ _ _ _. _. _

144 1 matter of fact, I don't believe that they did. I think their I 2 inference to me is that they did not ever find the correct 3 fuse before the valve went and closed by itself. 4 MR. REED: Thank you for that clarification. 5 MR. KERE: Do you feel better now about PORVs in 6 BWRs? 7 [ Laughter.3 8 MR. EBERSOLE: I think the next one now has to be oh, it 9 considered in the previous light of what happened at 10 was at Hatch also, wasn't it? 11 MR. JORDAN: No, this is the Oyster Creek. 12 MR. EBERSOLE: But there was an earlier one in which O k,,/ 13 we had these are scram dump volume leakage phenomena. 14 MR. JORDAN: This is Dave Powell, one of the 15 operations officers from ISE. 16 CSlide.] 17 MR. POWELL: Okay. Understanding that I have to be 18 a little bit brief on this, I'm going to go ahead and forget-19 about going over the event sequence, if that's okay. 20 I think there were quite a few members here last 21 time when I went over it, so if there are any questions 22 involving that, I will go ahead and entertain them. 23 MR. MOELLER: When it says that the plant is at 99 j () 24 percent power, was it steady state there? 25 MR. POWELL: Steady state, 99, t ,e m-.,, _ -. _,. _ -. -, - -. - ~ _. _, ~ _ _ _., _ _ _ _ _ _ _, -.. - _. __,m...---.__.,m-,

145 1 Okay, this event is of interest to us because it N.- 2 mimics an event that occurred at Hatch in August of 1982, 3 where they had a similar event which was an uncontrolled 4 leakage out of their scram discharge volume drain valves. 5 The reason that occurred was again they were unable 6 to clear the scram signal that they had in at that time which 7 was a high drywell pressure. And in this particular case 8 there were two valves in series. I will point them out right 9 here, these two. In the Hatch event, there was only the one. 10 This was a backfit modification, these two 11 particular valves. I'm not sure whether it was a generic 12 letter that required that, but I understand most of the BWR l \\ %/ 13 plants put this in at their first refueling outage following N 14 that particular requirement. 15 In this particular event following the scram, and in 16 the signal that was input, they were unable for 38 minutes to 17 clear the scram signal due to a 600 pound interlock which 18 seals in when they have an MSIV closure at this plant. That 19 is a rather plant-specific interlock. Most of the BWRs do not 20 have that. 21 These two particular valves failed. This particular 22 valve, they both had different failure mechanisms. This 23 particular valve failed to go fully' shut It went down to ( 24 about 1/8th of an inch from its seat. So there was about a 25 1/8th inch gap right there between the seat and the disk. _ ~ -

l ) 146 1 This particular valve is hypothetized to have 2 originally gone fully shut, but due to the pressure buildup 3 from this leaking valve, it forced this particular valve open. 4 Now the reason that occurred was they had an 5 improperly sized spring in the valve actuator. They had a 400 6 pound spring and it should have been -- well, I don't know 7 exactly what it should have been. 8 What they decided to do to fix it was to put in an 9 1100 pound spring. 10 Those two errors combined allowed an uncontrolled 11 leakage of reactor coolant. The estimate is about 500 gallons 12 that flowed from the scram discharge volume to the reactor i l' 3 building drain tank. From there it flashed. The fumes came 14 up through various vents in the reactor building at the 15 23-foot level, and the combination of the fumes from 'the paint and I'm not sure exactly where the 16 that was blistering 17 paint was. I believe the gentleman yesterday s&id it was i and in combination with the 18 upstream of the valves here 19 steam it ignited off the fire protection deluge system. And 20 as he indicated last time, it was on the 51-foot level, not 21 the 23-foot level 22 I think that lasted for approximately five minutes 23 before they got that secured. There was no damage to any 24 equipment inside the reactor building due to the actuation of 25 the deluge system or due to the steam. i

147 1 There was some radioactive contamination of the Us 2 23-foot level. Most of it was short-lived isotopes, and they 3 had to recover. 4 The potential significance for this is, of course, 5 the uncontrolled leakage of radioactive coolant outside of 6 containment. They also had, like I said, the fire deluge 7 system took off and actuated and, of course, any time you get I'm not sure if there was 8 water around electrical equipment j 9 any in that particular area, but any time you have the system 10 actuate, that potential is always there. 11 So it's a possibility at some other point, you know, 12 some other event, that that kind of a problem may show up, and 13 you might lose some secondary balance-of-plant or 14 safety-related equipment, 4 15 The third thing that occurred was we had CRD seal 16 high temperature alarms. They came in intermittently and the 17 problem with CRD high temperature alarms is not so much a 18 problem for this particular event. It is for later operation 4-19 of the plant. Should those seals degrade due to the high 20 temperatures, there's a possibility that they would be unable 21 to operate properly, either withdraw or insert properly or 22 possibly scram properly due to the degradation of the seals. I spoke 23 The two causes, as I noted earlier, was -- () 24 about these two particular failure mechanisms, I believe the 25 ultimate problem -- and I have to back up a little bit. There i 1

148 1 is no requirement to leak-rate-test these valves, all right? V,/~'g 2 These particular valves are categorized as B valves, Category 3 B valves under ASME Section 11 and because of that, these 4 particular valves were never leak-rate-tested after the 5 post-installation of the backfit mods, and had they done that these particular problems 6 -- you know, 20/20 hindsight 7 would have been probably picked up during the leak rate test. 8 Basically what they did when they put it in was just 9 hydrood the system to whatever the requirement was. To do 10 that, I would imagine they blank-flanged off the outlet valve, 11 opened up these two and just tested the piping within that 12 whole system as required under their ISI inspection program. \\ 13 MR. EBERSOLE: You had mentioned earlier, and I 14 think the Full Committee would like to know, that the 15 rationale for not testing these is that the minimum boundary 16 is consider ed to be at the rod seal itself, and all this junk 17 down here is just standard equipment, when in fact that's not 18 real life. 19 MR. POWELL: Well, I have the NUREG that addressed 20 that particular problem, and the Staff words on that 21 particular matter. Initially they came out and indicated that 22 the boundary should be at the scram discharge pilot valve, 23 which would be this, all right? This would still be outside 24 of that particular category A system, safety-related type 25 portion of it.

149 1 MR. EBERSOLE: Yes, but there's been a tremendous -~ V 2 controversy about potential leakage from the scram dump volume 3 even due to metallurgical failure of the volume proper, not to 4 mention valve failure as being a potentially serious event 5 which should have logically caused these valves to have been 6 recognized as safety-related valves. i ^ 7 MR. POWELL: Well, I'm not sure how each plant 8 chooses to look at their scram disoharge volume system, when 9 it comes to the requirements, but 10 MR. EBERSOLE: Isn't it a standard problem for all 11 BWRs? 12 MR. POWELL: I can't answer that. We haven't seen I 13 this type of problem, other than the Hatch event, which ! 14 guess precipitated putting on the series valves, hopefully to 15 preclude this kind of problem from occurring. 16 As far as the probability studies that were done on 17 the system, in terms of core melt sequence, this particular 18 NUREO addressed that also, and the conclusion was that failure and that's what they addressed, not the 19 of the piping 4 let's put it this would not lead to 20 valves at that time 21 way: 22 The probability was not high, compared with other 23 types of events that would lead to core damage. ( 24 MR. EBERSOLE: Oh, yes, the probability of piping 25 failure was virtually zero.

150 l 1 MR. POWELL: Well, they said it was less than 10 to .f-f 2 the minus 6 per plant year, so 3 MR. EBERSOLE: But they ignored the valves standing 4 right in the pipe. 5 MR. POWELL: Beg your pardon? 6 MR. EBERSOLE: They ignored the presence of the 7 valves which made an aperture through the pipe. 8 MR. POWELL: Okay. But I don't know. We're putting 9 out an information notice to show people that these kinds of and 10 events can occur. I would imagine that most utilities I should say the prudent 11 I would say probably mostly the will probably go ahead and change their IST 12 utilities 13 program to include those in a leak rate test on a yearly q 14 basis. 15 But again they are still not under any recuirement, 16 according, you know -- as*far as I know, to do that. 17 MR. EBERSOLE: The Staff will not upgrade these 18 valves? That's not 19 MR. POWELL: Well, I can't speak for them. I 20 understand that NRR is not really pursuing this matter much 21 more at this point, but they might. I don't know. 22 MR. JORDAN: There is not a decision yet to upgrade 23 the valves. The prompt action is to put out the information 24 notice and then make a subsequent decision on whether action 25 is needed.

151 1 l 1 MR. EBERSOLE: By the way, I don't believe you told ) \\ 2 us how they stopped this leakage. There's a little note 3 here. Did they try to blow down the system and get the 4 pressure down to reverse the interlook, and then did that 5 fail or what? 6 MR. POWELL: Well, as part of their normal 7 procedures for shutting down under MSIV closure, normally the 8 isolation condensers would have automatically set off for 9 them. The operators had to override that function because of 10 the high level that they had in the vessel at the time. 11 So what occurred was they got a high level with high 12 pressure and the electromagnetic valves, the A and D valves, 13 automatically actuated. 14 That basically blew down the system. It also, later 15 on in the event, they started up the reactor water clean-up 16 system and it began the letdown portion of that to the 17 condenser. That eventually resulted in some other problems. 18 I don't want to say problems, but they started getting 19 oscillations in the levels and they got other scram signals 20 that they picked up. 21 But, you know, I think that they had the plant they 22 pretty much under control. They just couldn't isolate 23 couldn't override the signal with the load switch at that time 24 until they got the pressure down below 600 pounds. 25 MR. JORDAN: The next event is the Rancho Seco

152 1 reactor coolant system high point vent leak. Howard Wong of O,' 2 IAE, who is one of the reactor construction engineers 3 MR. EBERSOLE: I think the Full Committee should 4 note that this is an event borderline to a one-inch small LOCA 5 on a B&W plant. 6 MR. JORDAN: It is appropriate that Howard Wong 7 should give the presentation. Number one, he was responsible 8 for the Bulletin 79-14, which was the "as-built" review, and 9 two, he went to Rancho Seco to assist Region V in follow-up of 10 this particular event, so he has physically been around this 11 piping and looked at it and followed upon the Licensee's 12 action. 13 MR. WONG: Let me set up the conditions that were 14 initially in the plant at the time of the event. The plant 15 was in hot shutdown. The correction on the slides says hot 16 standby was hot shutdown, the difference being that it was a 17 suboritical mode. We were starting up from a refueling 18 outage. 19 The event was a 20-ga11on per minute, non-isolable 20 primary coolant system leak on the high point vent system on 21 the B steam generator. 22 Initial identification of the leak was by personnel 23 who happened to be inside the containment who heard a pop 24 noise. On exiting containment, control room personnel were 25 informed, and by remote operated cameras inside containment, i L

153 f" 1 located the general location of the problem. r I l l 2 Control room personnel identified a small steam leak 3 using this camera. Additionally, the containment entry was 4 made to precisely locate the source of leakage, and it was l 5 identified as a non-isolable leak. At that point in time, the 6 normal cooldown procedure was ensued. 7 The actual leak portion occurred in a TMI 8 modification that was performed in the 1983 outage, which was 9 the high point vent lines. 10 What can be seen here is the RCS loop top of the 11 candy cane and the additional piping above was original 12 design. n \\~2 13 What occurred was a 120 degree through-wall crack 14 just below the weld. The enlarged diagram shows it, and 15 actually it was located on the outside portion of the T. This 16 weld was made during the 1983 outage. The old welds were this 17 piece in here and additionally up above. 18 The cause appears to be missing supports and fatigue 19 and therefore fatigue failure. 20 As a result of the RCS vent line addition, two 21 additional pipe supports had to be modified, and an addition 22 of one cross-brace member was required, Investigation 23 revealed that these support changes had not been performed, 24 although records stated to the fact that work had been done 25 and, indeed, inspected.

1 154 1 CBlide] \\ / i 2 I would like to show exactly what modification was 3 made, and this is perhaps a little more detailed than the 4 handout that is provided as an isolated sheet. The modification that was supposed to be made was 6 additional lateral bracing. Let me set it up first as the 7 piping system. Here is the RCS loop, the nossle coming off. 8 The TMI modification was this bottom line coming around. That 9 is the vent line. 10 The line that was the old piping was nitrogen supply 11 line, only used during refueling outages. 4 12 MR. KERR: Mr. Wong, that is an impressive diagram, O what are the salient points that 13 but what am I supposed to 14 I am supposed to learn from this presentation? 15 MR. WONG: The important points here are this is the 16 vent line that was added for TMI. Somewhat darker are the 17 modification work that was supposed to be made. The actual 18 break 19 MR. KERR: So what I should learn is that somebody 20 didn't do what they were supposed to do? 21 MR. WONG: That's correct. The supports that were 22 affected were these here, this cross-brace here, and also 23 notice this spool piece, removable spool piece in the middle. 24 MR. SIESS: Where was the crack? 25 MR. WONG: The orack is down in this portion here, _. _ -. ~ _ _ _ _ _, _ _ _. _ _ _ _.

=. 155 i right off the nozzle from the 'O 2 MR. SIESS: All that new stuff doesn't have any 3 braces on it? 4 MR. WONG: It does, but they are not shown on this 5 diagram. It does have pipe supports. 6 MR. SIESS: And this is braces on the old piping? 7 MR. WONG: That's right. This is more just to show 8 the pipe supports off of the existing system. It is important 9 to note 10 MR. SIESS: Even if this change hadn't been made, 11 there was a nossle there with pipe going into it before they 12 made it into a high point vent, right? I ( 13 MR. WONG: That's correct. If you take away this 14 portion, basically this is as was designed originally back in j 15 1974. 1 16 MR. SIESS: So that nossle would have failed or that 17 pipe would have failed even if this additional pipe hadn't 18 been put on it? 19 MR. WONG: Well, it's not so much additional pipe. and maybe I can jump ahead right 20 The point I want to get 21 now. The important is this removal of the spool piece in the i and this was to provide that the 22 middle. As designed 23 nitrogen system would not be contaminated during operation, so 24 it was planned to be removed during operation, during outages, 25 and would be in place for nitrogen blanket purposes. And if

156 1 that would have happened, the purpose of this was to transmit g-s 2 loads across both sides of the pipe. 3 The dummy spool piece was designed so that it would 4 go back during operations, basically giving rigidity to the 5 pipe, decreasing flexibility so that loads would be 6 transmitted across. 7 MR. KERR: Let me ask again what.I think Dr. Siess 8 was asking. Suppose that the TMI change had never 9 occurred. Would this leak have ocourred anyway? 10 MR. WONG: I think so. The fact of the matter is 11 that during operation, this piece was not here, basioally 12 leaving the one-inch pipe with about 60 pounds of distributed 0 x 13 weight, including a flange at this end, basically as a 14 cantilever. 15 MR. SIESS: It wasn't, the new piping that put the 16 stress on there; it was the old piping. 17 MR. WONG: It was most likely probably the fact that 18 this was not adequately supported. Without this spool piece, 19 it was hanging out ts a cantilever. 20 MR. SIESS-And that's the old pipe. 21 MR. WONG: That's correct. 22 MR. KERR: This is referred to as a high point vent l 23 leak. The high point vent was not even there before TMI-2. 24 MR. WONG: Well, it is an existing system as ( 25 established, as it was last month, the vent line was there. l . ~.... _ _.,...... - - -

157 1 MR. EBERSOLE: Yes, but originally it was not called 2 that. It was the nitrogen supply line. 3 MR. SIESS: But when did they take that spool piece 4 out? 5 MR. WONG: It looks like from investigation now -- 6 this was actually designed back in '74, this spool piece, to i 7 be removed during operations. 8 MR. SIESS: When did they take the spool piece out? 9 MR. WONG: They have always put it back during 10 refueling outages, so it appears from commercial operations 11 during an outage it was there, but during operation, it was 12 not, although a dummy piece should have been put there. 13, MR. SIESS: Now, was the weld that cracked in there 14 all that time subjected to the vibration and stress from the 15 spool piece? 16 MR. WONG: No, it was not. I will get back to the 17 original slide that I had of the weld. 18 MR. SIESS: Well, the original weld didn't fail; it 19 was some new one that was put in? 20 MR. WONG: That's right. This is the old weld. It's 21 a small pipe piece. Here is the T up in here. This weld here, 22 where the crack was initiated, it was a 1983 weld. 23 MR. SIESS: Well, why didn't the old one fail if it ) 24 was the old configuration that caused the vibration? 25 MR. WONG: There is a postulation that this is the . l

158 1 stainless to Inconel weld at this point, and this is stainless 2 to stainless. 3 MR. KERR: Okay. So it probably wouldn't have 4 occurred had the TMI-2 change not been made because you 5 wouldn't have had the stainless-to-stainless weld. 6 MR. WONG: That is one postulation, that's right. 7 MR. SIESS: You could think that way if you wanted 8 to. 9 MR. EBERSOLE: I think you are driving into that 10 point with some effort. 11 MR. WONG: The metallurgical examination of the 12 piping that had cracked showed evidence typical of high cycle

Bechtel, 13 fatigue.

The consultants which Rancho Seco used 14 GE, IMPEL and their own metallurgists, agree to that fact. 15 The primary crack is seen to be transgranular in an area of 16 high residual stress. Although it cannot determine the point 17 of crack initiation or the direction of propagation, evidence 18 characteristic of stress corrosion is missing. 19 MR. EBERSOLE: Well, what you are doing is going 20 through the physical details of why it failed, but I think the 21 real point of essence is did the paper records show that all 22 these pipes were in place and the hangers were in place and 23 all was in order? f) 24 MR. WONG: That is correct. U 25 MR. EBERSOLE: Well, the fact that that was not true

159 1 was the real root cause of this problem, wasn't it? 2 MR. WONG: That's correct. If it had been 3 implemented as designed, it would probably have not failed. 4 MR. SIESS: So if the paper says it is there, Jesse if the paper says it is not there, you know it's not 5 6 there. If the paper says it is there, it must be there. But 7 somehow we have got to tell that pipe 8 MR. EBERSOLE: That's right, we've got to tell the 9 pipe it's got to be there. 10 MR, SHEWMON: What sort of administrative action is 11 likely to be taken against the person who signed off on it 12 incorrectly? 13 MR. WONG: From the licensing standpoint, I can't 14 make a statement on that. I will get into corrective actions, 15 and that might clarify a little bit. Licensee actions, I 16 don't really know. 17 MR. SHEWMON: It is not a :nisdemeanor. 18 MR. WONG: I don't know. 19 MR. KERR: Well, that's QA. That doesn't have 20 anything to do with what is in the plant. 21 MR. WONG: It's a combination of not just QA/QC but 22 also the field engineer in this mode that had a package to be 23 sure it was properly done. The field engineer did sign off, 24 and then, of course, QC signed off. s 25 MR. SIESS: You mean QC signed off on the paper

fb '^ ? . t t l' t t-g 160 ~ 8-q ,z . f-- 1 without looking at it? The only person that is supposed to 'V s 25 look a t. it is the, field engineer, and everybody.else just c. 3 looks at paper? ,.f' + s3 g \\{. t ,3, ~. 'ik f'4 ~ 5 KR. WONG:. That i s ' n t *, correct. 'y f 5 MR. SIESS:9.' hope it isn't. It ain'tr,right. i,. t M MR. EB ER S O1.E : I understood that you could reach up t I, 7 with your hand and swi.sg this pipe back and forth'a foot or i t O two; is that right? 9 MR. WONG: I don' tDnow about that. It was cut up 10 by*the time'I got there. Dead weight analysis on this piece t \\' e

f 11 of pipe in tne "as-built" condition"shows a deflection of

.8 9", f y b.' e y g s 12 inches atrthe end of the piping r u n..; a t the end of t 'te flange, r-t i ( ,13 just on' dead weight alone. ) i 'I, s'o V \\ x s, o l 1 \\ 14" MR. S H E'W'40N : Wep1, b e f o r e z'y ou leave that one, wtere r oq 15 did the crack s.t. alt, outside or i n s i,d e ? ~ t s \\' f MR. WONG: On the ou ts i d e p'o r t i o n. 16 i ,e 17 MR. SHEwMON: Then stress, corrosion cracking isn't N <,. 7 < 18 r e C. l y germane.N \\ l'( ,7 ,f be $ 4 MR. WONG: It was on the outside. It appears >}o '[ 19 j v -l 8 I 2B N basically SHEWMONr }Well, )'t you have answered the question. MR. 21 ( 's e r '8 / 22 p, Go on. f J / g, f h 2f \\ N MR. WONG: An' additional point I want to make. \\ 24 There are axial indi c a t ions' along this full piece, t h e 's,w a l l it mi2ht be'ynty about three-quarters of,an i n c h.' 25 portion i . -. -, *..J_, ~ -., _

161 1 There are some slight axial indications. They are still j \\ 2 trying to investigate exactly what that is. They might be, 3 they consider, possibly from the manufacturing process. 4 MR. EBERSOLE: I think this is probably as far as 5 the Full Committee wants to hear this in light of the time 6 requirements here, unless I am wrong. Is the Full Committee 7 satisfied with this degree of presentation here? 8 MR. WARD: Yes. 9 MR. EFERSOLE: I think we can terminate this in the 10 interest of time and jump to the next one. 11 MR. JORDAN: The next one is Sequoyah Unit 2, and 12 MR. SIESS: I think the one we just heard about is a 13 QA problem that is much more interesting than the fact that an 14 overstressed pipe failed with fatigue, which is sort of nice 15 to know since that's the way we expect them to fail 16 MR. EBERSOLE: This is very brief here but very 17 pertinent, this one. 18 MR. WEISS: Sequoyah Unit 2 tripped on May 22nd from 19 100 percent power on overpower delta T. This event 20 demonstrates how a plant can trip following an approved 21 procedure, although a new one, despite all the precautions 22 that are in place to prevent maintenance or surveillance 23 activity from causing this sort of thing, such as O 24 communication between the control room and the people b 25 performing the surveillance or maintenance activity.

162 1 The particular surveillance or maintanance activity 2 under way at this time was the primary system calorimetrio 3 which was being performed for the first time. An instrument 4 technician had to take temperature readings from four 5 protection sets. The protection cabinets were located 6 approximately 15 feet apart, and he was using a digital 7 voltmeter to take the readings. 8 MR. LEWIS: Reading what? 9 MR. WEISS: Temperature, 10 MR. LEWIS: Well, what was he reading? j 11 MR. WEISS: He was supposed to be reading voltage; 12 and in fact, he had the leads connected to the digital \\s / 13 voltmeter'in the ammeter connections. 14 MR. LEWIS: Mr. Ebersole just said RTDs. Is that 15 correct? 16 MR. WEISS: Yes. l 17 MR. SHEWMON: But you missed his point. He had them 18 in the amp 19 MR. WEISS: So the internal resistance of the 20 voltmeter was much lower than it should have been. He had it 21 connected in the ammeter holes instead of the voltmeter 22 holes. So he tripped one channel, and since he is doing a 23 calorimetrio and he has to get a snapshot of the plant ) 24 conditions, he has to go to all four protection sets within 25 three minutes.

163 s 1 So he closes the door on one cabinet, moves to the J 2 next cabinet, opens that door, puts in the probes and takes a 3 reading. Although the reactor operator in the control room 4 noticed the dropping and noticed the trip, it happened too 5 quickly and the plant went down on two out of four 6 coincidence. 7 MR. SIESS: Well, that is improper use of test 8 equipment. You could have labeled it use of improper test 9 equipment just as well I mean if you checked out a voltmeter 10 and somebody was checking on the other end, he couldn't have 11 done this. So you say all precautions had been taken. I 12 think there is one I could have thought of. 13 MR. LEWIS: Well, you know, they blamed the other 14 voltage trip on the same thing. 15 MR. WEISS: Yes. You will recall there was an event 16 similar to this that we discussed before the Full Committee 17 some time ago where the use of a digital voltmeter caused a 18 shorting of the output transistors and the RPS system, and we 19 issued an information notice on the subject, The corrective 20 actions taken for this particular event include a precaution 21 about procedures not only regarding the consequences of making 22 a mistake but also for looking for the proper expected values 23 of voltage before proceeding to the next piece of equipment. () 24 MR. EBERSOLE: It would appear that TVA should 25 outlaw multipurpose meters.

164 /~N 1 MR. KERR: Well, there is also something to be said 2 for doing this sort of testing when the plant is not 3 operating, it seems to me. 4 MR. EBERSOLE: But this was a full power thermal 5 measurement, 6 MR. KERR: It doesn't have to be full power. It's a 7 temperature measurement, Jesse. That can be made at other 8 than full power. 9 MR. EBERSOLE: Well, this was a thermal heat 10 balance. They were on line. As a matter of fact, you remember 11 you said they had to do it within what? 12 MR. WEISS: Within three minutes. 13 MR. EBERSOLE: Three minutes channel to channel 14 MR. WEISS: Yes, and they were taking their readings 15 off of temperature resistance to voltage modules, they are 16 called. 17 MR. KERR: Well, you can surely take thermal 18 readings without voltmeters. ~19 MR. EBERSOLE: This is high and low temperature. 20 They are getting delta T. I think they had better outlaw 21 multi-meters. 22 MR. LEWIS: Well, or do less testing. 23 MR. SHEWMON: But Glenn will tell you these people O) 24 were selected maybe by not the right criteria and they should ( 25 be capable of knowing what a voltmeter is and what an ammeter i.

.. _ ~.. _. 165 1 is. 2 MR. EBERSOLE: It's the people who select the 3 meters. I 4 I think that does it. Thank you very much. 5 MR. WARD: Okay. We are finished with that. l 6 I suggest that the subcommittee that has to leave 7 could leave now. 8 MR. EBERSOLE: Thank you, Ed, and all your follows. 9 MR. JORDAN: By the way, t; ACRS has fed back to 10 the Staff the "thumby maintenance." We are using that word 11 now. 12 MR. WARD: This will complete the transcript for 13 today. i 14 CWhereupon, at 5:27 p.m. the reported meeting was 15 concluded.] 16. 17 18 19 20 21 22 23 24 25

1 CERTIFICATE OF OFFICleL REPORTER 2 3 4 5 This is to certify that'the attached proceedings 6 before the United States Nuclear Regulatory Commission in the 7 matter of. ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e 9 Name of proceeding: 303rd General Meeting 10 11 Docket No. DV) 12 place: Washington, D. C. 13 cate: Thursday, July 11, 1985 14 15 were held as herein appears and that this is the original 16 transcript thereof for the file of the United States Nuclear 17 Regulatory Commission. 19 (Signature) g (TypedName'ofReprter) /Suza$e B. Eedng 20 21 22 23 Ann Riley & Associates, Ltd. 24 25

HATCH UNIT 1 - STUCK OPEN SAFETY REl.IEF VALVE OF MAY 15, 1985 (G, RIVENBARK) A SYSTEMS INTERACTION EVENT UNIT 1 OPERATING AT FULL POWER CONTROL ROOM EMERGENCY VENTILATION SYSTEM CHARC0AL FILTER DELUGE VALVE ACTUATED WATER LEAKED THROUGH VENTILATION DUCTS'INTO A HATCH UNIT 1 ANALOG TRANSMITTER TRIP SYSTEM (ATTS) INSTRUMENT ' PANEL CAUSING SRV TO OPEN REACTOR MANUALLY SCRAMMED FEEDWATER PUMP REC 0 VERS REACTOR WATER LEVEL SRV CLOSED - WITHOUT OPERATOR ACTION CAUSE - LOSS OF lNSTRUMENT WATER SUPPLY CAUSING DELUGE VALVE TO OPEN TOGETHER WITH PLUGGED DRAINS NOT SURE HOW WATER CAUSED THE SRV TO OPEN ACTION REPLACED ATTS POWER SUPPLY, CLEANED PLUGGED DRAINS AND INSPECTED DRAINS IN REDUNDANT FILTER UNIT

~

LICENSEE PROPOSES TO ADD CLEAN 00T CHECK PROCEDURES FOR PLENUMS AND THEIR DRAINS ORAB WILL DEVELOP TIA TO COORDINATE: h IE NOTICE .;FURTHER INVESTIGATIVE EFFORTS GENERIC REVIEW 2

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1 0YSTER CREEK - UNCONTROLLED LEAKAGE OF REACTOR COOLANT /'~'T OUTSIDE CONTAINMENT X.) JUNE 12, 1985 (D. POWELL, IE) WITH REACTOR AT 99% POWER, FAILURE OF THE ELECTRIC PRESSURE REGULATOR CAUSED A TURBINE BYPASS VALVE TO OPEN RESULTING IN A REACTOR PRESSURE DECREASE, FOLLOWED.BY MSIV CLOSURE AND REACTOR SCRAM. SCRAM DISCHARGE VOLUME DRAIN VALVES FAILED TO FULLY SHUT / SEAT CAUSING REACTOR COOLANT TO BE DISCHARGED,TO THE REACTOR BUILDING DRAIN TANK. RELEASE OF' STEAM FROM FLOOR DRAINS ANb PAINT BLISTERING ON HOT PIPE CAUSES., PORTION OF REACTOR BUILDING DELUGE SYSTEM TO ACTIVATE I) SCPAM S'IGNAL NOT RESET F'OR 36 MIN ALLOWING CONTINUGUS %/ REACTOR COOLANT FLOW TO THE DRAIN TANK..CAUSE WAS 600 PSI INTERLOCK ON MSIV CLOSURE / LOSS OF CONDENSER VACUUM. SAFETY SIGNIFICANCE - (1) LOCK OUTSIDE CONTAINMENT, (2)-POTENTIAL EQUIPMENT MALFUNCTION DUE TO FIRE DELUGE SYSTEM, (3) EXCESSIVE CRD SEAL.TEMPERATUPES. CAUSE-VA,LVE SPRING ON VALVE V15-134 (VELAN) VALVE UNDERSIZED -VALVE V15-121 (VALTAK) STROKE DISTANCE INSUFFICIENT TO TIGHTLY SEAT THE VALVE. (1/8" OPENING) -IMPROPER POST-INSTALLATION TESTING OF VALVES ' CORRECTIVE ACTIONS - REPLACED 400 LB SPRING WITH 1100LB SPRING, ADJUSTED VALVE STROKE DISTANCE () CHECKED CRD' SEALS FOR DAMAGE, CHECKED ~.1 EQUIPMENT,' NO DAMAGE FOUND. NRC FOLLOWUP ACTION - IE NOTICE IN PREPARATION. 7

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DAVIS-BESSE - LOSS OF ALL MAIN FEEDWATER () AND AUXILIARY FEEDWATER JUNE 9, 1985 (A. DEAGAZIO, NRR) 1 Loss of one main feedwater pump at 90% power. Reactor trip at 78% power on high pressure. Both MSIVs close spuriously tripping remaining main feedwater pump. Steam generator low level starts both auxiliary feedwater -pumps but both trip on overspeed. 0perator erroneously manually trips SFRCS on low pressure No feedwater available for.about eight minutes and steam g-generator levels fell to about eight inches. PORV cycles three times - did not reseat on third cycle, ~ operators close block valve. Startup feedwater pump used to feed one steam generator. Operators restart auxiliary feedwater punps and restore normal post-trip conditions. No indication that subcooling margin was lost or that reactor coolant activity was abnormal. Plant now in cold shutdown. () 4 1 la

KNOWN EQUIPMENT FAILURES OR MALFUNCTIONS Main feedwater trip _ Spurious half trip of SFRCS MSIVEclosures (2) AFW pump trips on overspeed (2) AFW isolation valves' fail to open (2) PORY failure to reseat SUFP valve failure to open AFW speed' governor after reset Switchover to service water backup supply Damaged turbine bypass valve O e S O /3

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SSINS No.: 6835 IN 85-50 UNITED STATES V("N NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 July 8, 1985 IE INFORMATION NOTICE NO. 85-50: COMPLETE LOSS OF MAIN AND AUXILIARY FEEDWATER AT A PWR DESIGNED BY BABC0CK & WILC0X ADDRESSEES: All nuclear power facilities h'olding an operating license (OL):or construction pern.it (CP).

Purpose:

This information notice is being provided to inform licensees of a significant reactor operating event involving the loss of main and auxiliary feedwater at a pressurized water reactor. Information in this n6tice is preliminary and was obtained from the special NRC fact finding team which is investigating the event. A complete report of findings will form the basis for further communi-cations or actions related to this event. The NRC expects that recipients will review this notice for applicability to their facilities. Suggestions /l contained in.this notice do not con _stitute NRC requirements; therefore, no V specific action or written response is required. Description of Circumstances: On June 9, 1985, the Davis-Besse plant was operating at 90% power with Main Feedwater Pump 2 in manual control becaust problems in automatic had been experienced. A control problem with Main Feedwater Pump 1 occurred, and it tripped on overspeec. Reactor runback at 50% per minute toward 55% power was automatically initiated. Nevertheless, 30 seconds, later, the reactor tripped at 80% power on high pressure in the reactor coolant system. One second after reactor / turbine trip, one hannel of the Steam and Feedwater Rupture Control System (SFRCS) was automatically initiated due to a' spurious signal indicating low water level in Steam Generator 2. Both Main Steam Isolation Valves (MSIVs) closed. Three seconds after the actuation, the SFRCS' automatically reset. Closing of the MSIVs isolated the turbine of the operating main feedwater pump from its source of steam. The pump continued to supply feedwater to the steam generators for a few minutes as it coasted down. Four and.a half minutes after reactor trip, water level in the steam generators began 'to fall from the normal post-trip level which is 35 inches. After MSIV closure, steam release to atmosphere continued to remove decay heat. One minute later, Channel 1 of SFRCS actuated when the water level in Steam Generator 1 actually teached the SFRCS setpoint at 27 inches (See Figure 1). SFRCS started O Auxiliary Feedwater Pump 1 and initiated alignment of it to Steam Generator 1.' s

~ _ = d IN 85-50 July 8, 1985 Page 2.of 4 Within seconds after automatic initiation of Channel 1 of SFRCS, the operator actuated both channels of SFRCS; however,.he inadvertently actuated both SFRCS channels on low steam pressure instead of low water level. When an SFRCS channel.is. actuated on low steam. pressure, a rupture of the steam line asso'ciated i with that channel is presumed td have occurred. The SFRCS closes the steam generator isolation valves, including a valve in.the auxiliary feedwater line, and aligns the auxiliary feedwater pump to the other steam generator. Because both channels had been manually actuated on low steam pressure, both steam E . generators were isolated from both auxiliary feedwater pumps. Five seconds after the operator's zinadvertent actuation of both channels on low steam l pressure, SFRCS Channel 2 received an actual low water level actuation signal. Because low pressure initiation takes precedence, alignment of the auxiliary .feedwater pumps remained unchanged. At six minutes into the event as both auxiliary feedwater pumps were accelerating, they tripped on overspeed. ~ In summary, all main feedwater had been lost, both steam generators were isolated from feedwater and were boiling dry, all auxiliary feedwater pumps were tripped, i. -pressure of the reactor coolant system was rising, and reactor coolant system temperature was increasing. Within one minute after the operator's inadvertent actuation of the SFRCS on low steam presTure, the mistake had been recognized and the SFRCS had been reset. If equipment had performed in accordance with system design requirements, the operator's error might not have had a significant impact on the event. i e ' The auxiliary feedwater isolation valves should have reopened automatically, but the valves did not reopen. The operator then tried to reopen the valves from the main control panel, but the valves would not reopen. Operators were dispatched to locally start the auxiliary feedwater pumps, open the auxiliary e feedwater isolation valves, start the nonsafety-related motor-driven startup feedwater pump, and valve it to the system., Pres,sureandtemperatureinthereactorcIolantsystemcontinuedtorise because there was not sufficient water in the steam generators to provide an ~ adequate heat sink. At 13 minutes after reactor trip, reactor coolant system pressure reached 2425 psig, and the Pilot Operated Relief Valve (PORV) opened three times to limit the pressure rise. Ort the third lift, the valve remained .The operator closed the PORV block valve and reopened it two minutes open. later after the.PORV had closed. Approximately 16 to 18 minutes after reactor trip, the operators had the startup Water levels werd and auxiliary feedwater pumps running and the valves aligned. Reactor coolant temperature reached .beginning to rise in the steam generators. aihaximum of 594* F and then started to decrease to normal. Refilling of the, steam generators caused the reactor coolant system to fall to 1716 psig and about 540*F before returning to normal (See Figure 2). At 30 minutes after reactor trip, plant conditions were essentially stable. d.. e s /6

IN 85-50 July 8, 1985 Page 3 of 4 Discussion: For several minutes after reactor trip, the steam generators were unable to cool the reactor coolant system adequately. ~ The first problem contributing to this event was the loss of all main feedwater due to closure of the MSIVs. The licensee's hypothesis, based on information from Babcock & Wilcox, is that turbine trip caused a pressure transient upstream from the turbine stop valves which caused the outputs of the redundant steam generator level instrumentation channels to oscillate'widely for several seconds. The licensee believes that this caused a spurious low level actuation of SFRCS which closed the MSIVs. Three additional problems contributed to this event by affecting the availability of both trains of the auxiliary feedwater system. The first occurred when the reactor operator pressed the wrong SFRCS buttons. The second occurred when both auxiliary feedwater pumps tripped on overspeed. The third occurred when both auxiliary feedwater isolation valves did not reopen when SFRCS was reset. ~ Control buttons for the SFRCS are arranged in two vertical columns. Each column of buttons controls one SFRCS channel. The operator should have pressed the fourth button from the top in each column. Instead, the operator pressed the top buttons' causing isolation of both steam generators. /' Both auxiliary feedwater pumps are driven by Terry turbines which tripped on Q] overspeed early in the event. When this occurred, steam was being supplied to the turbines via crossover lines, which are longer than the normal supply lines and include long horizontal runs. The licensee believes that significant condensation may have occurred in the crossover lines. Further, the licensee believes that the quality of steam arriving at the turbines may have been affected significantly by the configuration of the crossover lines and may have caused the overspeed trips. The auxiliary feedwater system isolation valves have Limitorque motor operators. The motor operators have torque switches which prevent overtorquing of the valves by disconnecting power to the motors,. When the valves are being opened, additional torque is required to overcome friction while the gates are being unseated and while a significant pressure differential may exist across the gates. During the initial part of the opening stroke, the torque switch in the ' motor operator is bypassed by a bypass switch so that full motor torque is' developed if necessary. The licensee believes that these bypass switches went off bypass too early. The valves did not reopen until an operator unseated them by hand. e e. !7

IN 85-50 July 8, 1985 Page 4 of 4 ( ) No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate NRC regional office or this office. W n.A - award Jordan, Director ~ Divisio of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical

Contact:

R. W. Woodruff, IE (301) 492-4507 httachments: 1. Figure 1 - Steam Generator 1 Level and Pressure 2. Figure 2 - RCS Temperature and Pressure 3. List of Recently Issued IE Information Notices'. 4 , s _- s = w e e e e +, i O 4 j IN 85-50 l July 8,1985 Less ss 1 Su Ran:t LYL. t:3 (!m + ~ O 25 50 75 lod jg /@ /75 2oo L1r go (v 'PS32 SG 1 Oui STri PRESS.PT12B2 PSIR NO 6 50 70 0 750 800 SSD 900

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Attachment'3 IN 85-50 July 8, 1985 LIST OF RECENTLY ISSUED ,1(j IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-49 Relay Calibration Problem 7/1/85 All power reactor facilities holding an OL or CP 85-48 Respirator Users Notice: 6/19/85 All power reactor ~ Defective Self-Contained facilities holding Breathing Apparatus Air an,0L or CP, research, Cylinders and test reactor, fuel cycle and Priority 1 material licensees 85-47 Potential Effect Of Line-6/18/85 All power reactor Induced Vibration On Certain facilities holding Target Rock Solenoid-Operated an OL or CP ,yalves 85-46 Clarification Of Several 6/10/85 All power reactor 'z^s Aspects Of Removable Radio-facilities holding ) . active Surface Contamination an OL Limits -F.or Transport Packages 85-45 Potential Seismic Interaction 6/6/85 All power reactor Involving The Movable In-Core facilities holding Flux Mapping System Used In an OL or CP Westinghouse Designed Plants 85-44 Emergency Communication 5/30/85 All power reactor facilities holding System Monthly Test an OL 85-43 Radiography Events At Power 5/30/85 All power reactor Reactors. facilities holding. an OL or CP 85-42 Loose Phosphor In Panasonic 5/29/85 All power reactor 800 Series Badge Thermo-facilities holding luminescent Dosimeter (TLD) an OL or CP Elements [} OL = Operating License V CP = Construction Permit

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==oveoom.o co m -a west amas maan e.t aOCR,7 "== a m g,,,gg, ,,gg gg .... r .=.- = r Davis-Besse Unit 1 010 0 l3 0F 2 l2 h 0 11 }3 015 [o j o l0 l 314_f6 81 5 rum .- --=cs.-m.im 9 Description of occurrence: Davis-Besse Unit 1 was cperating at 90 percent cf full power with the No. 1 Main Feedvater Puup, MFP, operating in automatic and the No. 2 i MFP in manual control. This configu:ation was established to limit the susceptibility of the No. 2 MFP to control problems which had previously occurred. The control i problems occurred only af ter a reactor trip and appeared to be connected to the { automatic mode of operation. This configuration, therefore, permitted automatic feedvater control during operation and off ered improved availability of at least one t-MFP in the event of a reactor trtp. it 0135 hours, the No. IfMFP tripped en overspeed due to an unrelated cont;ol problem. The Control Roon operators increased the No. 2 MFP speed, but it did not 3 have adequate capacity, for the existing reactor power. The reactor tripped en high Reactor Coolant System, RCS, pressure at 0135:30 hours, tripping the turbine. Reactor power was at approximately 80 percent of full power at the time of the trip. 4 T ~ Inmediately following the trip, a spurious Steam and Feedvater Rupture Control System, SFRCS. lov stes:n generator level full trip occurred on Channel 2, an SYRCS 4 ? full trip alatu vas received, and both sain steam isolation valves, ESlVs, closed. An actual lW steam generator level did not exist at this time. This s Nrious trip g resulted in a partial actuation of the STRCS components since caly the MSlVs actuated. -i L h!hentheXSIVsclosed,themainsteamsupplywasisolatedtotheMFPs. The No. 2 7 NFP continued to supply feedvater until approximately 0140 hours at which time its discharge prsssure was not high enough to supply feedvater to the stesa generators. The level in the steam generators.which was bein5 enintained at the low level limit setpoint (35 inches) began to decrease. SFRCS Actuation Channel No. I then automati-- cally initiated on lov steam generator level, starting the No.1 Auxiliary Feedvater s Pus.p, AFP, to feed the No. 1 Steam Generator (see Attachment 1 for a diagram of I SFRCS actuated cenponents). 5 At 0141:08 hours, a Control loom cperator attempted to manually initiate the SFRCS, g h however, he incorrectly actuated the SFRCS on low steam pressure instead of the 3 3 desired lov steam generstar level. Therefore, each STRCS actuatien channel sensed f that its respective stars generator was depressurized. STRCS Actuation Channel [ Ho.1 then attempted to align ATP No. I to feed Steam Generator No. 2. STRCS Actuation Channel No. 2 attes:pted to align ATP No.' 2* to.faed Steam Generator No.1. Both actuation channale closed their respective Auxiliary Feedvater Containment Isolation Valves (AF399,~ AF60PJ, which prevested any auxiliary feedvater. flow from T reaching the steam generators. At 0141:31 hours ATP No. 1 tripped on overspeed, j At 0141:44 hours, AFP No. 2 tripped on overspeed, g i At 0142:00 h& Irs, an operator recognized the manual initiation error and reset the -) E low pressure STRCS buttons, and pushed the lov steam generator level STRCS manual g s'ctuation buttons. Since both SFRCS actuation channels were already tripped on lov 3 steam generator level, the SFRCS automatically began to realign the AFPs when the low pressdre buttons were reset. However, the Auxiliary Feedvater Containment [ { Isolation Valves (AF599, AF608) did not automatically open. The operators attempted g; E to open these valves from the Control Roon by operating their control switches and f { by reinitializing the SFRCS. These attempts failed to open the valves. Equipment 6 a E g._, g

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w,e m Davis-Besse Unit 1 o l5 lo lo l0 !3 !G [6 9!5 01!!3 Ol0 Ol a OF 212 mr u ww s ama.,im Operators were sent to open these valves locally, and when the valves were moved off their closed seats' utilizing the manual handvheels, the motor operator responded and fully opened the valves. During this period, attempts were also being made to restart the ATPs and preparations were undetway to start the motor operated Startup Teodvater Pump. The RCS average temperature was increasing due to the lack of primary to secondary heat transfer. RCS pressure was increasing due to th,e decreasing density of the RCS vater and increasing pressurizer level,. RCS pressure increased to the Power Operated Relief Valve, PORV, setpoint (2425 psig): The PORY cycled a total of three times, relieving pressurizer pressure to the Quench' Tank. Following the third opening. the PORV failed to reclose at the proper RCS pressure. The Control Room operator observed the primary plant conditions and closed the block valve on the FORV. ICS pressure was at approximately 2075 psig when the block valve closed. The Quench Tank contained the discharges from the PORV. At approximately 0151 hours, the operators placed the Startup Feedvater Pump in operation to supply the steam generators. Steam Generator No. I pressure had decreased to appreximately 750 psig. Steam Generator No.1 repressurized to approxi-mately 900 psig from the Startup Feedvater Pump. Steam Generator No. 2 had decreased l lto920psig. At 0152 tours, the No. 2 AFP was returned to service by the operators locally, Maximum RCS temperature had reached approximately 592 degrees Fahrenheit. At 0155 hours, the No. 1 ATP was returned to service locally by the operators. Control of the AFP turbines was maintained locally by an operator at the turbine trip throttle valve. At 0158 hours RCS average temperature was rascored to the normal post trip temperature. The cooldown of the RCS lowered RCS pressure to a mini =um of approminately 1720 psig. Operators manually started the No. 1 High Pressure Injection. HPI, Pump in the piggyback mode (Decay Heat Pun:p No.1 supplying the suction to the EPI Pump No.1) in precautionary anticipation of the rapid cooldown. Only a slight amount of water (less than 50 gallons) needed to be injected. Stveral other equipment malfunctions occurred which did not affect the physical plant response. One source range nuclear instrumentation, NI, channel was inoperable prior to the trip. The remainbg source range NI channel failed to indicate properly j vhen'it was automatically energized after the trip. The display units far.tha S:fety Parameter Display System, SPDS, were inoperable in the control Roem at the time of the trip. At 0158:40 hours, the suction of the No. 1 ATP automatically transferred.from the condensate Storage Tank, CST, to the Service Water System. The cperstar manually realigned the pump suction back to the CST. No significant amount cf service water was added to the steam generator during the recovery from the transient. It was noticed that t% pneumatic operator on one main turbine bypass valve was damaged, preventing the valve from being opened. This did not affect the post transient response of the plant. Additional-details of the plant transient and corrective actions vill be provided in the restart report response to the Region III Confirmatory Action 1.etter (85-06). Attactment 2 provides a chronological listing of the event. This report is being i submitted in compliance with paragraph 50.73(a)(2)(1). 50.73(a)(2)(iv), 50.73(a)(2)(v), [ g.u. -

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.... n= = Davis-Besse Unit 1 d l0 0l5 0F 2 l2 o p lo j o je l3 tt.16 815 "' ll 3 rm u-. 4 --ac s amewm 50.73(a)(2)(vi), and 50.73(a)(2)(v11). This report also satisfies the reporting requirements for a Taergency Core Cooling System Actuation Special Report, Section 6.9.2(a) of Technical Specifications. This was the fourth high pressure injection actuation cycle _to date. Designation of Apparent Cause of Occurrence: This transient was initiated when the No.1 MFP developed control problems and tripped on overspeed. The plant tripped on high RCS pressure due to inadequate feedvatar being supplied from the No. 2. MFP turing the plant u nback. The cause of the MFP overspeed tripping was defernined to l be due to a bad speed sue:r.ation and valve lift reference circuit board card in the -MTP control. A' frequency to voltage converter chip had failed. The board is being ret'urned to Ceneral Electric for further analysis on the root cause of the failure. Iba root cause of the MSIV closure has not yet been detetuined. It is pr'esently believed that the MSIVs properly responded to a monentary low level SFRCS trip. Further investigations vill follow once an action plan is completed. The cause of the STRCS spurious trip on low stean generator level has not yet been positively determined. Troubleshooting vill begin in accordance with the action However,it is presently believed that the steem generator level sensing l plan. l channels are sensing an extremely rapid secondary side pressure transient that cecurs in the steam generator following the turbine stop valve closure on a turbine trip. These level transmitters share a cosmon set of sensing lines with transmitters which were replaced during the 1984 Refueling Outage. Prior to the 1984 Refueling Outage, Bailey BY level transmitters were installed which have nov been replaced by Rosement Model 1153. Since these Rosecost transmitters have no significant displace-mest required for operation, while the Bailey BYs required a volune displacenent to operate the bellows, it is postulated that. the responsiveness of the sensing line and transmitter arrangenent has been greatly increased by this change. This increased responsiveness allowed the STRCS to sense,the rapid secondary side pressure transients which previously were undetected. Furth u analysis of this condition is undervay. The cause of the incorrect manual SFRCS initiation was personnel error attributed to a poor svitch layout. These STRCS canual initiation pushbuttons had been identified in the Detailed Control Room Design Review as one of the principal items needing human engineering improvenants, n ere are two adjacent vertical columns of buttons vith five buttons in each column (see Attachment 3 for arrangenent details). Each column represents one SFRCS actuation channel. To manually initiate both channels of the,SFECS for steam generator low level, the operator shculd have depressed the fourth button from the top in each colunn; instead, the two top buttons were depressed. A design change had been developed prior to this event *co improve the switch layout and vill be' implemented during this outage. The cause of the ATPs tripping on overspeed af ter initiation has not yet been positively determined. Water flashing through the nozzles of the AFP turbines is thought to be a contributor. The governor was inspected on both ATP turbines, and no contributing factors to the overspeed were seen. Further investigations and testing are planned. g..

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=====ne a a' ua===en 1 -==a m m:. m 0;vis-Besse Unit 1 06 or 2[2 o p;o;olo13 l4 6 8 ! 5 0 l13 0 l0 1 9 1 a u- . <,6 me m me wm m The cause of the Auxiliary Feedvater Yalves (AF599 and AT608) not opening by the motor operator was determined to be a combination of a high differential pressure and an improperly set torque switch bypass limit switch. V,ith this torque switch bypass limit switch improperly sat, the motor operator was allowed to torque out during the opening stroke. nese valves are open during normal operation and were closed by the incorrect manual initiation of the STRCS. If the ATPs had been cperating at the time the valves attempted to open, the differential pressure across the valves would have been significantly lover, and the valves should have opened to allow the auxiliary feedvater flow to occur. These valves were stroked following the transient an.1 ability of the valves to open (without significant differential-pressure) was verified. Recent testing also verified that the valve operators torque out under high differential pressure with the improperly set torque switch bypass limit switch. Further investigations are in progress. The cause of the control problems with the ATPs after the overspeed was reset is presently attributed to the difficulty in opening the trip throttle vslves. No mechanical deficiencies were found while investigating the resetting of the overspeed trip device'/ linkage. Further investigations are in progress. { }ne cause of the PORY not properly ressating has not yet been positively identified. Operator observations at the time of the transient indicate that the electronic controls signal was calling for the valve to reclose. A visual inspection and disasse=bly of the PORY failed to identify the cause. Further investigations are in i

progress, The two, independent SPDS display units were inoperable due to separate but similar failures in the data transmission system between the Control Roem terminals and their respective processors. The failures are of an inter =ittent nature and the exact cause is still under investigation.

ne cause of the source range NIs inoperability has not been positively identified, ne failure of the source range N1s has been a repetitive proble: at Davis-Besse with repeated investigations failing to determine the root cause. Since 1977, the boron trifluoride detectors, preamp, and cable in Containment have been replaced, f,..a}qng with the modules in the Resctor Protection System and a reworking of the grounding on the preamp and count rate amplifier module connections.* No positive effect on the total eltatnation of tha spiking, nor the erreneous/ elevated count rate has occurred from these corrective actions. Further review is being performed on the-possibility of ground loops, induced current or voltage from adjacent cables, or intermittent problems with the count rate amplifier ecdule. The cause of the turbine Eypass Valve 2-2 damage has not been identified, ne valve was disassembled and the actuator stem extension piece was found bent, four parts were missing, and the valve internals were found loose. Several valve parts were shipped co the.endor for further analysis. Further review of the turbine bypass v l Ivalvefailureisunderway. %P.( NOW W S4$'

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om .m .... r-Mg- = D vis-Besse Unit I o is lo is lo 13 tb 0 l 113 O!0 o l7 7 2b 8 15 -w. -. --c w ana m The cause of the inadvertent ATP No. I suction supply transfer'ftca the Condensate Storage Tank to Service Water has not yet been determined. Testing and other investigstions are currently being performed. The cause of MS-106 apparently cycling in about one third of the expected stroked tir.e is still under investigation. Analysis of Occurrence: This event involved a temporary loss of feedvater to the This event was bounded by the analyses previously performed (see

steam generators.

Toledo Edison submittals to NP.C Serial No. 506 dated May 22, 1979, and Serial No. 517 dated June 15,1979). which analyzed a loss of all feedvater for 30 minutes following a reactor trip. These analyses showed that as long as either 1) Auxiliary Feedvater is' restored within 30 minutes of the loss of main feedvater. OR 2) Vithin 10 minutes, at least one makeup pump an( the PORY are available for { ) primary cooling.(feed and bleed) and the Startup Feedvater Pu=p is avail-able to supply a steam generator, fuel cladding temperatures would remain within a few degrees of saturated fluid temperature and no cladding rupture or metal vater reaction would occur. Operator interviews indicated that the shift was fully aware of the core status and vare prepared to implement the " feed and bleed" core cooling method if the auxiliary fcedvater was not restored. The Startup Feedvater Pump was available throughout the event and in fact was placed in service within ten minutes of the tripping of the AFPs. Auxiliary Teedvater was restored within 12 minutes of the loss of feedvater. These response cines and equipment availability are vell within the loss of feedvater analyses. At no time during the event was the required subcooled margin (20 degrees Fahren-heit) lost. The peacgr coolant pumps continued ~to operate throughout the event. The primary code safety valves were not challenged and at no time during the event did toe ECS pressure or temperature exceed the allowable values. The maximum temperature reached was below the normal operating temperature for the hot leg There is no indication of any fuel cladding degradation based on the temperature. rcactor coolant radioche=istry analysis. An analysis has been performed by Babcock & Wilcox to determine if the transient ~ adversely aff ected the stem generators. Conditions and couponents specifically analyzed include: (1) Main Teedvater Nozzles, (2) Auxiliary Feedvater Rozzles. (3) Steas Gene'rator Tubes. (4) Tube to Shell Delta T's, and (5) Lover Tubesheets l I The results shev that the transient had no adverse structural effect on the steam generators. j g...

c.... b w UCENSEE EVENT REPORT (LER) TEXT CONTINUATION wie one =o 2, 4,w n=u me - maans its secast - e m W we m l 4W q=eg;,wg Davis-Besse Unit ! 011 D OL O OB 0F 1 !2 o l5 l0 } o l01314 5 915 we .ca m..<m Corrective Action: ne failed circuit board vill be replaced in the No.1 MFP. As a precautionary measure, the No. 2 MFP speed control circuit vill also be inspected for a similar failure. A ne corrective action on the MSIV closure has not yet been determined since trouble-shooting has not yet begun. The corrective actions for the STRCS spurious trip on low steam generator level have ost yet been determined since troubleshooting has not yet begdn. L e proper method of manual

actuation of the SFRCS buttons vill be reviewed with all licensed operators.

He 'svitch layout is being modified to add additional deserkation of the actuation buttons, and to add actuation guards over the switches (see Attachment 3). D e corrective actions to be taken to prevent the ATP trip on overspeed have not yet b:en determined. The torque switch bypass limit switch will be reset on the Auxiliary Feedvater Valves AF599 and AF608. Maintenance personnel vill receive additional instruction, and the procedure for setting the motor operator valve limit switches will receive { haditional clarification. other nuclear safety related motor operated valves at Davis-Besse vill be evaluated. The corrective actions to correct the control probleza with the AFPs af ter the overspeed was reset have not been identified. Corrective actions to be taken on the POR7 have not yet been identified. Corrective' actions for the repair of the data transmission systems aff ecting the SPDS Control Room displays have not yet been identified. n e corrective actions for repair of the source range N1a have not yet been determined. The pneuma:ic actuator for Turbine Bypass Valve 2-2 vill be replaced. Additional corre4.tive actions may be necessary af ter further invastigation to determine the rcot cause of the fat. led valve actuat6er >- A tabulation of the causes and corrective actions deter.ained to date is suur.arized in Attachmetit 4. ,, Corrective action details f or the No.1 AFP suction supply transfer from the Conden-cate Storage Tank to Service Water has not yet been identified. Corrective action details for MS-106 have not yet been identified. Further investi-g: tion is in progress. I hailureData: This is the first occurrence at Davis-Besse of a loss of both nain and auxiliary feedvater. g ew m.

1 ~.2 = - - = - = = c -- - - - - -m-- ,.m.% __ 7 es.mausesmem - e w maa UCENSEE EVENT REPORT (LER) TEXT CONTINUATION ne.= oves me se no.co takett Lif s cacarv -_--an as tse - A se l .aet se , maan ses jw ;p p = :- Davis-Besse Unit 1 3 l5 l0 {o l0 f3 h l6,8 5 01113 0l0 01 9 0F 2l2 rea a . m m a=c m aa.v a This is the first f ailure that has occurred at Davis-Besse on the MFP turbine alcctronic controllers which has caused sa overspeed tripping of the pumps. A new electronic control system for main feedvater pumps was installed during the 1986 Refueling Outage. Spurious closures of the MSIVs have occurred previously.at Davis-Besse before time dalays were added to the steam to f eedvater pressure differential trip circuitry. An STP.CS spurious half trip on low stein generator level has occurred on two 'previ-cu's trips since the 198'4 -Refueling Outage. Spurious trips on los steam generator level have not occurred prior'to the 1984 Refueling Outage. Incorrect manual initiation of the STRCS has not previously occurred at Davis-Besse. The AT?s tripping on overspeed af ter initiation has not previously occurred at Davis-Bessa. The Auxiliary Feedvater Valves AF599 and AF608 are norr. ally open. One previous occurrence of one of these valves not opening with high dif ferential pressure ccurred af ter the March 2.1984 reactor trip. n a operators do not normally attempt to control the AFP turbines locally. Problema with centrolling these pumps do not appear to have been repetitive, however, some probleca have been experienced previously with proper resetting of the trip throttle valve. The PCRV has not oeen challenged since the pressure setpoint was raised in 1979. Prior to 1979, several deficiencies were noted in the valve operation. In September 1977, the valve stuck in the open position, causing an overpressurization of the quench tank. l The diversity of the SPDS display sources (Ramtak and Chronatics display devices) has not= ally allowed at least one SPDS display to renain overable, ne failure rate of these units is higher than is acceptable. Efforts are underway to increase the l system reliability. g, ne f ailures of the source range NIs have been a repetitive occurrence at Davis-Besse l even though exhaustive evaluations and corrective actions have been taken. ThedamEgedpneunatieoperatorontheturbinebypassvalvehasnotpreviously occurred at Davis 3Besse. l There have been several cases where the AFP suction inadvertently transferred fron the condensate' Storage Tank to the Service Water supply. ? i I I I Rioort Nor__ NF-33-SS-18 D71 No(s): 85-088-g,.

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