ML20138A673
| ML20138A673 | |
| Person / Time | |
|---|---|
| Issue date: | 03/14/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1497, NUDOCS 8603200018 | |
| Download: ML20138A673 (375) | |
Text
Y/4W ORJGINAL D UN11ED STATES NUCLEAR REGULATORY COMMISSION IN THE MATTER OF: DOCKET NO:
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 311TH GENERIsL MEETING LOCATION: UASHINGTON, D. C. PAGES: 1 224
! DATE: FRIDAY, MARCH 14, 1986 m
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PUBLIC NOTICE BY THE
-UNITED STATES NUCLEAR REGULATORY COMMISSIONERS' l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l
I FRIDAY, MARCH 14, 1986 W
The contents of this stenographic transcript of the proceedings of the United States Nuclear Regulatory Commission's Advisory Committee on Reactor Safeguards (ACRS), as reported herein, is an uncorrected record of the discussions recorded at the meeting held on the above
- date.
No member of the ACRS Staff and no participant at this meeting accepts any responsibility for errors or inaccuracies of statement or data contained in this transcript.
l
CR26178.0 DAV,SIMONS 1 Oj9 o I g UNITED STA'11'S OF AMERICA
)
2 NUCLEAR REGULATORY COMMISSION 3
4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 311Til GENERAL MEETING 5
6 Nuclear Regulatory Commission Room 1046 7 1717 H Street, N.W.
Washington, D. C.
8l Friday, March 14, 1986 9
10 The 311th General Meeting reconvend at 11:40 a.m.,
11 ; Dr. Paul G. Shewmon. presiding.
J l
ACRS MEMBERS PRESENT: j C)
' 13 '
I7 DR. WILLIAM KERR 18 DR. IIAROLD W. LEWIS 10 DR. CARSON MARK 20 MR. CARLYLE MICIIELSON DR. DADE W. MOELLER 21 * ,
1 i l DR. DAVID OKRENT DR. PAUL G. SilEWMON !
22 h j MR. GLENN A. REED DR. CilESTER P. SIESS t 23 0 !
p DR. FORREST J. REMICK MR. CilARLES J. WYLIE v un i Ac2 Faired Flemrters, Inc.
25 , j l
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l l 1780 01 01 2 q
V DAVbw 1 PROCEEDINGS 2 DR. SHEWMON: The Chairman's Report of ,
3 Modification, GDC 4 allowed applications, the leak report l 4 break concept and all high energy lines for nuclear power l
l 5 plants. People familiar with how steel pipes fail have 6 always known that a sudden double-ended guillotine break is
~
7 essentially incredible; however, it seems like a 8 conservative bounding assumption for seismic containments.
9 So the NRC promulgated it and the industry learns to live 10 with it; however, as people, the NRC and the industry got to i
11 taking this idea literally. Massive constraints to cope 12 with massive forces. For example, pipe whip restraints and 13 snubbers to handle arbitrary breaks in the primary system.
14 Pipe whip restraints and snubbers in balance of plant ECCS 15 systems to handle the consequences of large sudden breaks. ,
16 Looking for a more rational yet conservative 17 approach for research on elastic plastic failure or elastic 1 18 plastic fracture mechanics presented a means of justifying 19 more reasonable alternatives. This work has shown how to 20 calculate when a through wall crack in a high energy line 21 would be stable instead of leading to a sudden failure.
22 That is, a leak before break. The formalism has been 23 well-proven as a means for determining how fast the crack 24 will grown in the structural members and when it becomes i O 25 unstable.
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(- DAVbw f-s) 1 The technology has been identified for many years 2 in aircraft and large structures like offshore drill rigs.
3 The acceptance by the NRC of the concept of leak before 4 break has led to a series of changes. The first application 5 came in the resolution of USIA to assymetrical blowdown, 6 where it was determined that the primary piping of the power 7 plant would leak well before it broke, and as a result, one 8 didn' t have to rebuild the primary to resist assymetric 9 blowdown loads.
10 The next application was to the pipe whip 11 restraints in the primary system. This was embodied in the 12 so-called limited scope revision of GDC 4 that we reviewed a 13 few months ago. This has gone out for public comment.
14 What is before us today is the so-called " broad 15 scope revision" of GDC 4 that CRGR has approved and the 16 Staf f is ready to have sent out for public comment. This 17 would allow the application of leak before break to lines in 18 the balance of plant and admit severe criterias such as a 19 low failure probability and are expected to exhibit an 20 easily detectable weak weld before fast failure.
21 Details of this criteria will be developed in a 22 Reg Guide which has yet to be written.
23 The Subcommittee met on the 27th and 28th.
24 Harold Etherington was there and Dave Ward, Mike Bender, Ed 25 Rotaba and Tom Kasner as consultants. The representatives ACE-FEDERAL REPORTERS, INC.
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(_) DAVbw 1 of Beaver Valley 2 came in and talked to us. This is the 2 lead plant in trying to apply the concept of leak before 4 3 break for the entire plant.
4 EPRI came in to voice their interest in this and 5 talk about their research program concerning it. j l
6 Westinghouse gave a presentation as an interested j 7 vendor. Sid Bernstein, representing AIF, was there, in 8 essence saying, we think you should have done this years !
4 9 ago. l 1
10 These items and other things are summarized in i
11 the minutes of the meeting, which are in your notebook. We 12 recommend approval of issuing the rule for public comment, b
v 13 and a letter has been drafted to this effect and is I
14 available within a planned presentation on what this change 3
15 would be, and Staff available to answer questions.
16 Yes?
i 17 MR. EBERSOLE: Is there any differentiation I
18 between the guillotine break and the linear split, such as t 19 happened in the steam plant recently, in the context of
- 20 leaking?
i I
21 DR. SilEWMON: In the context of what I had to say 22 for experience, or what?
! 23 MR. EBERSOLE: I find it easier to understand i
24 that you would deny that the guillotine break will occur i
25 than I do the linear split.
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k_) DAVbw 1 DR. SHEWMON: As you know, primary systems are 2 not built with longitudinal welds, and the steam lines that 3 opened up were longitudinal welds. We will be getting into 4 steam lines in nuclear plants which are big enough so that 5 they of ten do have longitudinal welds. My impression is 6 that the best thinking on the steam exerciso versus nuclear 7 plants is that they are operating in a range where creep is 8 important and those produced effects that you can't detect 9 with NDE, the way you can with a fatigue crack.
10 In fact, the suggestion of the consultants and 11 the ready agreement of the Staff we have in our letter that 12 the application of this will be limited to lines of 650F, J
13 which basically would exclude the gas-cooled reactors, at 14 this time at least or the application up to that.
15 So partly there aren' t any longitudinal wells, 16 but there are some, and the feeling is that the ones that 17 fail have been more a factor of the temperature they were 18 at, and we wouldn't exclude that.
19 DR. OKRENT: It seems to me that approving the 20 general idea without knowing what the specific restrictive 21 constraints are going to be is, in this case, a bit 22 premature for me, because what could be allowed, might, in i 23 fact, encompass in the end, all of the existing LWRs. I J
fs 24 have no way of knowing. I think, as I understand the 25 practice in Germany, for example, they do this sort of thing ACE-FEDERAL REPORTERS, INC.
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_a DAVbw 1 on what I'll call a fotDard fitting, where they demand high 2 quality in the fabrication and insist on changed methods of 3 fabrication, et cetera. And I can understand that# approach 4 and would really not have much trouble going along with it.
5 But I think, in fact, what one is talking about here, are 6 essentially all the plants that are built, with maybe a few 7 exceptions, and I myself have a problem with what I will 8 call a " pig in a poke."
9 DR. SHEWMON: Well, I think there's a couple of 10 things I did not mention here and should have, that 11 everybody who han looked at it, I think, agrees that there 12 are some real benefits in taking out restraints that are 13 being talked about. It would ease inspection. It would 14 make the lines more flexible for temperature change and 15 would probably be an improvement of safety, if they were 16 taken out.
17 Another aspect of it is that these lines have 18 been quite reliable, but where they have not been reliable, 19 they would be excluded, and I think of water hammer as one 20 example of that and corro'sion as another.
21 And third, what the rule would do would be to 22 allow the Staff to establish criteria for the application of 23 this, and my own concern is that the Staff is likely to come 24 up with such onerous requirements on the criteria for O 25 applying this, that the won' t take out as many restraints ace FEDERAL REPORTERS, INC.
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i 1780 01 06 7 DAVbw 1 as they probably should.
l 2 But that is something which will be evolved and i
4 3 something which we certainly will be allowed to comment on.
j 4 So it seems to me, if we vote against the rule at t
5 this point, it is more voting against the concept of tr.ying 6 to take out some of these restraints.
I I 7 DR. OKRENT: I'd like to continue the i
8 discussion.
i j 9 It is my impression that the way the Germans did
) 10 it is, they decided what would be good enough, so that they i
, 11 wouldn't need to consider breaks, and then they said, okay,
, 12 for plants built this way, we will go along that way. That v 13 is not what we are proposing to do here.
, 14 Now I didn' t say I would oppose any relaxation, t
- 15 you know,'of the existing situation. I am not arguing that i
! 16 the current restraints necessarily end up on the plus sidor i
j 17 however, I am -- at the moment, I have no basis for being i
i 18 convinced that whatever it is that is dono for plants of f
l 19 what I will call lessor quality than what the Germans are i
j 20 requiring and perhaps in some casos, relatively unknown
! 21 quality, that in the first place, there exists a means of l 22 monitoring which would accomplish, let's say, a good 23 substitute for original quality. In other words, that you
! 24 can, in fact, inspect meaningfully for the kinds of things
!( 25 you are concerned about, and you know what the philosophy is i
i
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j 1780 01 07 8 DAVbw 1 and what we will expect.
) 2 Secondly, with regard to water hammer, I have not l 3 seen anything that tells us which pipes are not going to be 4 subjected to water hammer in what I will call complex 5 transients, of which we have one at the San onofre 1. The l 6 Staff.has chosen to look at the water hammer question, in f 7 terms of what has happened in the past, not what may happen, 8 and I haave no feel for what role this can play in-the f 9 overall matter. I have heard people say that you can
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i 10 develop water hammers large enough to rupture pipes. I am i
j 11 not an expert in water hammer. So again, I have a feeling j 12 that I am being asked prematurely to be in favor of l' 13 something that I might well be agreeable to, if I saw that 14 the things we've done --
1 15 DR. SHEWMON: tMll, you haven't heard the j 16 presentation. If you feel this way, I wish you could have
! 17 attended the meeting, where you would have gotten a chance 1
l 18 to do this. You are right, the Germans are doing this in a I
] 19 forward-fitting arrangement. I don't think you are right i 20 with regard to whether we monitor and with regard to l
1 21 inferring things about the poor quality. These were built 1
22 to code. There are special tests that people can do and 23 will do and have done on these materials. So I think we 24 know it better than you think we know it.
O 25 MR. REED: Let me understand this rule.
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DAVbw 1 DR. SHEWMON: We will have a Staff presentation 2 that will explain this.
4
) 3 MR. MICHELSON: Is that today?
4 DR. SHEWMON: Right as soon as we finish asking 5 questions.
i i
6 MR. MICHELSON: We can ask the Sta,ff? .
I i 7 DR. SIIEWMON: That's right.
1
- 8 MR. REED
- I have to be worried about some kinds 1
9 of pipes.
j 10 DR. SHEWMON: The Staff shares your concern about n
4
,R 11 some kind of pipes.
1 12 Do you have any comments, Dave?
- 13 MR. WARD
- No, I don't. !
! 14 DR. SHEWMON: liarold?
j 15 MR. ETHERINGTON: In-service inspection is
- 16 required of pipes. We will not approve use of degraded j 17 piping. So we have some continuing surveillance.
. 18 DR. OKRENT: I agree, but what constitutes an j
19 adequate surveillance? The British use the term "a
{
f 20 validated surveillance."
l 21 MR. ETHERINGTON: A surveillance which discovers ;
! 22 that nothing is required by the Code. !
l l 23 DR. SilEWMON : Did you have a comment? l i
i 24 MR. MICHELSON: I have some general comments, 1 C:)
l 25 which I would like the Staf f to think about during their i
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- discussion. .I assume they will clarify exactly what types 4
l 2 of breaks will still be postulated outside of containment, a
! 3 if any. I would like to know what kind of breaks. I am J
4 quite interested in the design of subcompartments and
- i 5 compartments relative to the pressure produced by the
- 6 previously postulated breaks and what will now be the design 7 basis for subcompartments. I am a little worried about ,
I 8 walls not being retianed in an appropriate condition, i 4
) 9 DR. SHEWMON: You're talking about future plants? j 10 MR. MICHELSON: Present plants. If I have to go
! 11 in and cut a hole in the wall to do something, will I have l
12 to put the hole back in its original state? If there is no 13 other breaks in that hole, perhaps not. But then I have to l
14 ask the pipes of things, besides pipes that can break, valve I
15 flanges, valve bonnets, pump internals. These things --
l 16 corrosion, molding, whatever, can lead to large breaks, 17 which do not follow the leak before break kind of rules at I -
i 18 all or may not. They don't behave like pipes. They are i
4 l 19 totally different. Bellows are also different.
- i j 20 DR. SHEWMON: In pressure bolting f ailures, we
- 21 had a pretty good history of leak before break.
1 1 22 MR. MICHELSON: Depending on who you talk to, i
23 whether they zipper open or not. That is a very good 1 24 question. 1
(:) 25 DR. SHEWMON: Do you have any examples?
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(_) DAVbw 1 MR. MICHELSON: No. We fortunately haven' t had 2 these large breaks yet.
3 DR. SHEWMON: We:also have a lot of bolts.
4 % MR. MICHELSON: Well, they don't follow your nice i 5 fracture mechanic theories. It is unrelated to it. So 6 there h s got to be some new theory to give any comfort that 7 flanges will never come off.
8 DR.' SHEWMON: You have been given other theory in 9 which people talk about the number of excess bolts they have 10 and the probability of all of them failing at once by 11 corrosion process or whatever it is.
12 MR. MICHELSON: But that is another whole t
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13 presentation, if that is included as a part of leak before 14 break, and I think you have to ectablish that on an equally ,
15 firm basis, as you do the fracture mechanic postulations of 16 these other leaks. So that is another area I would like to 17 hear a little more about.
18 And there is another thing coming, and that is
.. 19- , the decoupling of SSEs and LOCAs, and I would like to hear 20 the Staf f's current position on whether or not you get LOCAs 21 during SSEs, since it is apparently a part of our proposed 22 letter, as well, this draft letter that you wrote.
23 DR. SHEWMON: That has nothing to do with the 24 evaluation of ECCS in this.
> , i 25 MR. MICHELSON: Excuse me. Maybe it was back ACE FEDERAL REPORTERS, INC.
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s1 DAVbw 1 in your minutes, where it was discussed. I am sorry; you 2 are right. Back in the minutes, it talks about decoupling 3 SSEs and LOCAs as being acceptable. I would like to hear 4 the Staf f's current thoughts on that. I guess it was 5 discussed during the Subcommittee?
4 6 DR. SHEWMON: Staff's position earlier was that 7 they would not allow leak before breaks. They would get 8 lines which exhibited stress corrosion cracking. My
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9 impression is that on the current generation, they would not
- s 10 allow this, if these were not considered subject to stress 11 corrosion cracking.
12 (Slide.)
13 MR. O'BRIEN: Mr. Igne is now distributing a copy 14 of the broad scope rulemaking package, plus a copy of my 15 slides. I would like to direct Dave Okrent immediately to 16 my acceptance criteria. I was told I only had half an hour 17 today, so I didn't bring all my slides.
18 DR. SHEWMON: We don't want to know everything 19 you know. We will have some questions.
20 MR. O'BRIEN: This first slide indicates our 21 schedule of the limited scope rule. We already have four of 22 the five Commissioners voting af firmatively without comment 23 They were supposed to affirm their vote yesterday, but 24 because of eleventh hour questions from Commissioner 25 Asselstine on backfit, they didn' t af firm it. We are ACE-FEDERAL REPORTERS, INC.
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(Q -DAVbw 1 hoping that we can get affirmation of the limited scope rule 2 next week, and we wish to propose the rule in the Federal 3 Register in May of this year and the broader rule in 4 December of '86, in time to assist the efforts of Beaver 5 Valley, which is due to go on line in February of '87.
6 MR. MICHELSON: Limited scope means inside 7 containment?
8 MR. O'BRIEN: No. It means primary coolant 9 quality.
10 (Slide.)
11 My first three slides deal with issues that were 12 raised at the Subcommittee meeting on the 27th. We d id n' t 13 resolve thme the way you recommended, but the first question 14 is, is the rule limited only to high energy piping? We said
- 15 it numerous times, and we took the trouble now to define 16 high energy piping, which is somewhere buried in the Federal 17 Register list that was just handed out to you. I believe it 18 was five or six or seven. We did have a definition of high 19 energy piping. Then there were some discussions as to where 20 the rule applies to specific break locations or to the 21 entire system. It was always our intent to apply the rule 4
22 to entire fluid system piping or portions between enchors, i 23 which are analyzable portions of the analysis.
24 1
25 ACE-FEDERAL REPORTERS, INC.
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(_/ DAVbur 1 We stated this more emphatically somewhere in the l
2 Federal Register notice, somewhere on page 7. i l
3 MR. EBERSOLE: Shouldn't there be some qualifier l 4 that the pipe has to be not subjectable to abnormally high 1
5 pressures, such as coupling from LOCA, or that it in fact is !
l 6 free fron the influences of water hammer? l l
7 MR. O'BRIEN: Yes, that is in our acceptance 8 criteria. There is a discussion of that in the rule.
9 MR. EBERSOLE: For instance, the low pressure 10 piping that we might connect to the high pressure panel 11 system, if you connect it all bets are off.
12 MR. O'BRIEN: You mean accidentally?
(~x
\- 13 MR. EBERSOLE: Yes.
14 MR. O'BRIEN: No, we don' t require anybody doing 15 anything beyond the design basis.
16 MR. EBERSOLE: Of course, the argument is you 17 will never do that.
18 DR. SHEWMON: John, I wish you would start with 19 the rule instead of answering questions which the people 20 here don' t know anything about. Maybe they will come up 21 later.
22 MR. O'BRIEN: That means we can skip three 23 slides?
f, 24 DR. SHEWMON: You may for now, but I think you k_)
25 had better set the stage. I mean, you would do better to ACE-FEDERAL REPORTERS, INC.
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1780 02 02 15 DAVbur 1 set the stage.
2 (Slide.)
3 MR. O'BRIEN: That means that I begin with this 4 slide, right? Okay.
5 The first three slides will be deferred, I 6 gather, till the end, and instead we begin a discussion of 7 the rule.
8 I have a feeling I made a bad choice not to 9 discuss the acceptance criteria this morning, but the issue 10 is that we have --
11 DR. SHEWMON: You will discuss the acceptance 12 criteria before we get done?
13 MR. O'BRIEN: Without the benefit of slides, 14 yes.
15 We have about 15,000 pipe whip restraints, 16 costing from 2- to $6 billion, depending on how you reckon 17 costs, with an average life of around seven years each, 18 giving us 100,000 experience years with not even one single 19 instance where the pipe whip restraint was needed.
20 There have been pressure boundary failures due to 21 water hammer at Indian Point and Maine Yankee, but they did 22 not require pipe whip restraints. We had a failure at Duane l
23 Arnold as well, a very large crack. !
1 24
-s So we spent a lot of money for something we have 25 never needed, and when we look overseas we have never found ACE-FEDERAL REPORTERS, INC.
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(; DAVbur anybody who needed it.
_ 1 2 We feel that piping integrity can be degraded by 3 pipe whip restraints if they inadvertently restrain thermal 4 growth, which could happen either by having failed 5 snubbers. Failed snubbers means that you have got a bad 6 thermal analysis, which means you don' t know how the pipe is 7 growing, and a failed snubber can lead to contact between 8 the pipe and the pipe whip restraint, which could grow a 9 crack at a location not now protected by pipe whip 10 restraints.
11 (Slide.)
12 DR. OKRENT: Excuse me. Has that occurred, to
(/
N- 13 your knowledge?
14 MR. O'BRIEN: No.
15 DR. OKRENT: That is a hypothesis?
16 MR. O'BRIEN: That is a hypothesis.
17 DR. OKRENT: I agree it could occur, but I am 18 just trying to point out that you are saying there is no 19 experience of large pipe breaks. I was just trying to say 20 you also don' t have this particular experience.
21 DR. SHEWMON: We have torn things out of walls.
22 DR. OKRENT: Water hammer?
23 DR. SHEWMON: No. I mean with thermal g_ 24 expansion. But that certainly is overloaded.
V 25 DR. OKRENT: That I can't say. But I am not ACE-FEDERAL REPORTERS, INC.
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2 We feel that piping integrity can be degraded by 3 pipe whip restraints if they inadvertently restrain thermal 4 growth, which could happen either by having failed 5 snubbers. Failed snubbers means that you have got a bad 6 thermal analysis, which means you don' t know how the pipe is 7 growing, and a failed snubber can lead to contact between 8 the pipe and the pipe whip restraint, which could grow a 9 crack at. a location not now protected by pipe whip 10 restraints.
11 ( Sl ide . )
12 DR. OKRENT: Excuse me. Has that occurred, to 13 your knowledge?
14 MR. O'BRIEN: No.
15 DR. OKRENT: That is a hypothesis?
16 MR. O'BRIEN: That is a hypothesis.
17 DR. OKRENT: I agree it could occur, but I am 18 just trying to point out that you are saying there is no 19 experience of large pipe breaks. I was just trying to say 20 you also don' t have this particular experience.
21 DR. SHEWMON: We have torn things out of walls.
22 DR. OKRENT: Water hammer?
l 23 DR. SHEWMON: No. I mean with thermal 24 expansion. But that certainly is overloaded.
25 DR. OKRENT: That I can't say. But I am not ACE-FEDERAL REPORTERS, INC.
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(_) DAVbur 1l aware -- I mean I don't recall cracks being identiried in 2 pipes. I know Nine Mile Point had a problem, but that was 3 due to design error.
4 MR. O'BRIEN: Anyway, continuing on why pipe whip 5 restraints may contact piping, besidas failed snubbers we 6 have problems with support gaps, which are very difficult to 7 deal with in thermal analysis and also difficulty in 8 retaining close tolerances and alignments of very large 9 structures.
10 Some of these pipe whip restraints weigh 30 to 40 11 tons and have to be placed correctly. They have to be 12 placed in cold condition, and when the piping is hot they I l
(~b 13 have to be in some cases a fraction of an inch but more 14 typically an inch or two away from the pipe.
15 Dan Landis tells the story of a pipe predicted to 16 move four inches to the right. But it was observed to move 17 two inches to the left.
18 It is very difficult to do a good thermal 19 analysis of piping when you get it with gaps.
20 Apart from this, pipe whip restraints will 21 probably foul up the seismic performance of piping because 22 there is a possibility of impact loads between the pipe and 23 the pipe whip restraints during an earthquake and obviously 7_ ,, 24 the diminished effectiveness of in-service inspections.
U 25 For all of these reasons, we feel that pipe whip ACE-FEDERAL REPORTERS, INC.
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(_) DAVbur 1 restraints have had a negative safety impact.
2 (Slide.)
3 In our rule we limit ourselves only to dynamic 4 effects, which means that the double-ended pipe rupture is 5 still postulated for EQ. It is postulated for containment 6 des ig n , and it is postulated for ECCS performance.
7 MR. MICHELSON: You are talking here about a 8 broad scope rule. Let's talk about outside containment.
9 What are you postulating?
10 MR. O'BRIEN: At this particular moment, there is 11 no relaxation of any requirement dealing with EO.
12 MR. MICHELSON: How about subcompartment design?
(')
\/ 13 MR. O'BRIEN: Pressurization of subcompartments?
14 MR. MICHELSON: For instance --
15 MR. O'BRIEN: That is eliminated except if that 16 subcompartment is part of the containment volume.
17 MR. MICHELSON: Why did you eliminate the 18 subcompartment design, which has nothing to do with all the 19 good arguments about getting rid of the restraints?
20 MR. BOSTAK: It is a dynamic effect.
21 MR. MICHELSON: But what has it got to do with 22 the problem of too many restraints, and so forth?
23 MR. BOSTAK: It is a dynamic effect, and if you 24 are eliminating dynamic effects, it is not a leakage crack
/s i l
V 25 type effect. And the other thing you want to keep in mind l
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\_1 DAVbur. 1 is these postulated break locations are at finite spots, and 2 they are not everywhere in the plant.
3 So in these other areas you have nothing. You 4 haie no EQ yualification. You have nothing at all. So in 5 other words, it is a mild environment.
6 MR. MICHELSON: I don' t exactly agree with that.
7 It depends on what part of the plant you are talking about.
8 MR. BOSTAK: But that is what it is if there is 9 no postulated accident, and you have to keep that in mind.
10 MR. MICHELSON: But there are postulated breaks 11 of steam lines outside of containment?
12 MR. BOSTAK: There are, but they are at finite 13 locations. They are not everywhere.
14 MR. MICHELSON: That is right, but they are in 15 rooms that then have to be designed fcr the effects of those 16 breaks?
17 MR. BOSTAK: That is right, but that is a dynamic 18 effect, and if you have gone through the fracture mechanics 19 evaluation, again it is on these lines.
20 MR. MICHE LSON: Let me back up from a different 21 l direction. If you have already spent your money and you
, 22 have already designed and built the compartment, why are you 23 now saying*it no longer has to be kept in that condition?
. 24 There is nobody complaining about it to begin 25 with. You are giving away a safety. feature.
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1780 0 2 07 20 DAVbur 1 MR. BOSTAK: But we are not telling anybody that ,
2 they need to change it one way or the other. If at'some 3 point in the future a person came in and discussed this 4 point that you are bringing up, I guess we would look at it 1
5 on its merits.
6 MR.-MICHELSON: I am not even cure it is any 7 longer an unresolved issue. If I have a compartment in 8 which there are now no postulated breaks, I don' t have to 9' retain the pressure features of that compartment.
j 10 MR. BOSTAK: That is correct.
11 MR. MICHELSON: Why do I need to come and ask the 12 NRC? If I have to cut a hole in it to come and get the 4
13 piece of equipment out, I will leave the hole in it. I 14 don' t have to ask the NRC.
i
- 15 MR. BOSTAK
- That is exactly correct.
16 MR. MICHELSON: You have given the future away l
17 then.
18 .DR. SHEWMON: But it is not clear to many of us 19 that it was indeed a safety feature to begin with in terms
! 20 of reducing risk to the health and safety of the public.
21 MR. MICHELSON: Well, yes, it was because what 22 happened is a lot of people would use stacked blocks, for l
23 instance, for some of these walls. They have had to now go 1
24 back. They had to in the past go back and reinforce them O 25 because they found that even on modest pressure rises the i
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(_) DAVbur 1 stacked blocks would come out like buckshot.
2 Now, you are saying that you no longer have to do 3 this.
4 DR. SHEWMON: This is only when you postulated 5 this guillotine break.
6 MR. MICHELSON: It doesn' t even take that big a 7 break necessarily. Yes, it is from the guillotine break.
8 DR. SHEWMON: It seems to me what they are doing 9 with regard to other breaks, in my opinion, is a fair 10 question. But coping with this hypothetical instantaneous 11 double-ended break is just nonsense, and to call it a safety 12 feature offends me in several ways.
k- 13 MR. MICHELSON: Well, it bothers me a little bit 14 though to degrade the level of safety we have already built 15 in.
16 DR. SHEWMON: The alleged degree of safety?
17 MR. MICHELSON: Okay.
18 MR. O'BRIEN: The point that has been made is 19 that the static pressures are still there. It is the 20 dynamic effects, the transients, the short-lived --
21 MR. MICHELSON: I haven't heard the first 22 economic justification for the degrading of suboptimum 23 design conditions because it is already there. Nobody is 7- 24 asking anything to be changed.
V 25 DR. SHEWMON: Let me come back. You say the ACE-FEDERAL REPORTERS, INC.
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1780 02 09 22 DAVbur i pressure is, say, a slow leak.
2 MR. O'BRIEN: It pressurized the compartment, but 3 there will be no differential pressure across the wall.
4 What we are eliminating is the short-lived, rapid 5 pressurization that vanishes. We have to exhaust a 6 tremendous amount in the compartment in a short time. That 7 only happens on a very rapid break.
8 MR. MICHELSON: You haven' t defined any 9 intermediate size of break. You have now said it is leak 10 before break. There are no breaks at all.
11 You have removed the requirement that you are 12 confining in many ways --
13 MR. O'BRIEN: Dynamic means time variant.
14 Dynamic effects are eliminated. The pressurization of the 15 compartment from a leak of a valve or whatever, a seaI, 16 those are still there.
17 MR. MICHELSON: That is the part that is not 18 clear.
19 MR. O'BRIEN: Static means static, and dynamic i
20 means dynamic. We are wrestling about words.
4 21 DR. SHEWMON: They have been wrestling with it 22 for a little while then, because your point is that if we l l
23 have a big leak that causes nothing to whip but it releases l 24 a gosh awful amount of steam into that chamber, anything l b
a 25 having to do with pressurization of that chamber that was 4 <
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l ) DAVbur' 1 there before is still there under the new rule.
l 2 MR. O' BRIEN: I think he is affirming that.
I 3 MR. MICHELSON:- That is not my understanding.
4 But if it, that is great. I withdraw my objections.
1 5 MR. WICKMAN: The rule only deals with piping. !
6 We do not deal with pumps. We do not deal with valves. We 4
j 7 do not deal with anything else except the. piping.
I 8 MR. MICHELSON: That is another thing that gives i
, 9 me a problem. We won' t get into it yet. Remember the 10 bellows, and so forth.
L" I 11 MR. ETHERINGTON: This is going to come up
. 12 later.
< ( 13 MR. WICKMAN: I know of no bellows-in high energy .
14 lines, and if there are, they would merit special staff
, 15 consideration.
16 MR. EBERSOLE: May I ask a question which may I
1 17 clarify?
18 What are we getting , and would this rule j ust i,
19 apply --
i F 20 DR. SHEWMON: Carl keeps wondering. I don' t know 1
i 21 whether he has a question he would like to have the staff f
22 answer or not, t
23 MR. MICHELSON: I think they have answered the
- 24 question. They are taking care of the pressurization fix, i
(:) 25 just not the dynamic impact.
1
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-DAVbur 1 MR. WARD: No, you are missing the point now.
2 They are taking care of pressurization of the whole 3 containment, but not of individual compartments because that 4 is a dynamic effect.
- 5 MR. O'BRIEN
- If it is the differential pressure 6 that worries you, when you pressurize one side of a wall, f
7 the other side not, in the long term, because most 8 compartments are not sealed, then you get the pressurization
! 9 on both sides. What we are eliminating is this short-lived l 10 dynamic ef fects that happen from very rapid pressurization
! 11 from slow leaking from the rest of the building.
12 MR. MICHELSON: I understand the concept, but I 13 don' t find it in the material that I read on A-46, but maybe 14 I missed it. I don' t find it.
1 15 MR. ETHERINGTON: It was just a jet effect, t
16 wasn't it?
i 17 MR. O'BRIEN: No. I have my next slide. It says 18 pressurization of compartments is one of the dynamic f' 19 effects.
20 DR. OKRENT: He says that is eliminated, i 21 pressurization of compartmengp.
i 22 MR. WARD: From individual compartments.
23 DR. OKRENT: Individual compartments. l 4
24 MR. MICHELSON: Outside of containment.
25 MR. O'BRIEN: May I continue, or should I )
~ l I
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-g C/ DAVbur 1 continue? j l
DR. SHEWMON:
2 I think so. Go ahead.
3 MR. O'BRIENs And then leak before break. The 4 reason we didn' t apply it to containments is because we have 5 to have a replacement pipe rupture.
6 When we apply leak before break, there are no 7 dynamic effects associated with pipe ruptere in that
- 8 particular system. It is reduced to zero. We have no 9 replacement pipe rupture.
10 MR. MICHELSON: Dynamic effects, including 11 pressurization?
12 MR. O'BRIEN: Of compartments. We didn' t involve 6
i,.
13 ourselves in containment of ECCS and EQ because that would 14 slow down this action, and we wanted to get -- we saw a
15 immediate benefits, and we didn' t want to delay achieving 16 these benefits for a number of NTOLs which are coming .
i 17 online.
18 We acknowledged that we have to do this at some
, 19 time in the future but not now.
20 DR. OKRENT: Excuse me. This would again allow I 21 steam lines to run right alongside control rooms because the 22 worst you could have is a leak before break?
23 MR. O'BRIEN: If you really believe --
24 DR. OKRENT: The staff would have no legal basis 25 for saying no, I believe.
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DAVbur 1 MR. BOSTAK: I think the staff does have a basis 1 2 for arrangement, and we do that now. If you go back in i 3 time, this whole thing started with respect to a steam line 4 through a control room.
5 The staff probably I think was mistakenly 6 prompted into postulating these mechanistic breaks, but we I 7 have the authority. We have the ability to tell people to 4 8 move equipment, move lines if they are in the vicinity of 9 things like control rooms.
10 DR. OKRENT: What is the authority you have to do 11 this?
12 You said it, but I am trying to understand what T
' v 13 the basis is if the only thing you are allowed to say is 3
j 14 that this can have a leak before break.
I 15 MR. BOSTAK: If you have' vital equipment, 16 equipment required for shutdown, control room, any of those T
17 spaces, and the fact that you do have a high energy line 18 that is in the vicinity of that, you don' t have to postulate 19 a break in order to get the utility to either move the 20 equipment or move the line.
21 DR. SHEWMON: You are saying that will be in your 22 criteria or what?
23 MR. BOSTAK: That is already in the standard 24 review plans on separation. This is the first thing that 25 you try to do. You try to achieve separation. You don' t ACE-FEDERAL REPORTERS, INC.
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1780 02 14 27 ;
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. DAVbur 1 have to go through and postulate a break to get somebody to 2 move something.
'3 DR. OKRENT: I am sorry, I am not talking about 4 the steam lines going through the control room because I 5 think for environmenta'l reasons you might well be able to 4 6 say how are you going to environmentally qualify anything.
7 But alongside it is not clear to me, but outside a
8 it is not clear to me on what basis you would say no.
i 9 10 11 12 13 i 14 15 d
I
?
i 16 i
17 i 18 19 20 21 22 23 24 -
25
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ss DAVbur 1 MR. BOSTAK: If you are really concerned about 2 that space -- and there are some you would be concerned 3 about -- you go through and call it nonmechanistic 4 evaluation of what the effects of the failure of that line 5 would be.
6 By the way, all of the accident evaluations that 7 depend on postulated breaks are still there. There is no i
8 change. This is the Chapter 15 accident evaluations. They 9 are all still there.
10 DR. OKRENT: If the members think they understand 11 where this will all lead, good. I don't.
12 DR. SHEWMON: Go ahead.
13 MR. O'BRIEN: The final reason we limit ourselves 14 to dynamic effects is that it has, we believe, a negative 1
15 effect on plant performance, and we thought we could address 16 the most pressing issue first.
17 ( Sl ide . )
18 Here is the definition of dynamic ef fects. Pipe 19 whip, jet impingement, and decompression waves. This will
- 20 have an effect on the way reactor internals are designed in 21 the future, steam generator internals.
22 We do allow the dynamic effects associated with 23 pressurization in cavity subcompartments. We need this, by 1 24 the way, to get out of A2. We are already receiving a 25 lesson on this for A2.
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(_) DAVbur 1 And finally, any missiles associated with pipe 2 rupture.
3 (Slide.)
4 We did a little survey recently in Columbus.
5 Actually, we had an international leak before break 6 seminar. We had participation from a number of nations.
7 A couple of general notes. Everybody applying 8 leak before break will not deal with containment ECCS or 9 environmental qualification and will have some kind of 10 criterion for leakage detection.
11 The United Kingdom are intending to reject leak 12 before break because of what we think are unrealistic
,~~
13 concerns about stress corrosion cracking and NDE, which is 14 probably a more realistic concern.
15 The French are also inclined to reject leak 16 before break but are pursuing the issue.
17 The Germans are very advanced in their 18 application of leak before break.
19 The Italians are close to the Germans.
20 DR. OKRENT: I think the record should show, 21 though, that it is in a front-fitting, not a backfitting i
22 way. j 23 MR. BOSTAK: If I could comment j ust briefly on 24 that. I was at the seminar, and we talked to many of the 1
(
25 people. In fact, we put together a slide, and on the older ACE-FEDERAL REPORTERS, INC.
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1780 03 03 30 s DAVbur 1 plants there may be some breakdown in communications, in the 2 language. But I think if you went into these plants, you 3 wouldn' t find the number of pipe whip restraints, jet 4 impingement shields that we have.
5 In other words,-they may have said that they are !
6 taking care of postulated pipe breaks, but when you get 7 right down into the plants, you will probably find that they 1
8 are not. So they don't have the problem on the old plants i
9 that we have because they don' t have any of these things 10 there.
11 DR. OKRENT: And they are not removing anything.
12 They are just demanding higher quality and better built 13 primary and secondary systems, and they are not removing 14 constraints because they weren't there.
15 Is that what you are saying?
16 MR. BOSTAK: I am saying to a large degree, as 17 best we can gather from talking to people'that have been 18 through the plants, they don' t have the numbers of pipe whip 19 restraints and jet impingement barriers that we have, that ,
20 you would expect if they were really following the same 21 criteria that you have.
22 DR. OKRENT: That is a little different. I have 23 no idea what numbers they have, but it was my impression 24 that in fact they were making a change that did involve --
25 MR. BOSTAK: On the convoy plants the Germans .
I 1
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- d DAVbur 1 are.
2 DR. SHEWMON: Everybody agrees with you on that.
3 Let me hold up for one minute, Jess, to see what the other 4 comment was here. j i
5 MR. MAYFIELD: Mike Mayfield, from Research.
6 Until just recently, we would have agreed with 7 your suggestion that the Germans were looking only in a 8 forward design capacity. However, in a meeting February 3rd 9 with Professor Kusmal from the NPA and in a paper that was 10 just received, published by a guy from Kraf twerk Union, they 11 are backfitting to the boilers. They have an analysis 12 scheme. We are finding that they are repiping plants. It
, 13 ' is not something that they planned to do. It is in present 14 and past tense.
15 DR. SHEWMON: They have repiped boilers once.
16 Are they repiping them a second time?
17 MR. MAYPIELD: Apparently, this is relatively 18 recent kinds of changes to the boilers. If they can show by 19 analysis that they can meet a leak before break, fracture 20 i resistant -- the terminology gets a little poorly defined --
21 but they are allowing leak before break concepts to be used 22 .in a backfitting mode.
23 MR. MICHELSON: Outside of containment as well?
24 MR. MAYFIELD: I don' t think so.
O. 25 MR. MICHELSON: I keep hearing two things, you ACE-FEDERAL REPORTERS, INC.
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1780 03 05 32 k_m) DAVbur 1 know, one about inside, one about outside containment. We 2 are talking broad scope, and you say the Germans are doing 3 it.
4 You should make it clear when you are talking 5 about how they are treating outside.
6 MR. BOSTAK: On the convoy plants they are going 7 outside.
8 DR. SHEWMON: It is not clear how we are treating 9 outside of containment, at least, because the staff said 10 they would need some kind of leak detection out there. Yet 11 there is no sumps. It is not clear how the industry would 12 come back in and say they had good leak detection on those
(~)/
13 lines.
14 My impression of the staff position is that they 15 would require that.
16 MR. MICHELSON: I was wondering if the Germans 17 had decided to go up on scope in theirs.
18 MR. REED: I think that there is a slide coming 19 up, hopefully the second one from this one, that will 20 clarify a lot of things, and maybe it should even come up 21 first because I have been sitting here in a state of f rig ht 22 ever since I read this proposed rule allows application of 23 leak before break technology to all high energy, 275 psi or 24 200 F fluid system piping to demonstrate that specific pipe 25 ruptures need not be treated on the design basis.
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/ DAVbur 1 Now, that frightens me. But if I get to this 2 slide, I get unfrightened.
- 3 DR. L'..EWMON
- Fine. Let's let him get on then.
t 4 Perhaps we should shut up and let him go on, but
- 5 we won't. Go on.
6 MR. EBERSOLE: Let me ask just kind of a rude .
7 . question. We have had some spectacular pipe breaks which 8 weren' t expected.
9 DR. SHEWMON: If it is for information on what he 10 is saying -- if it is a hypothetical, let's let him finish.
11 MR. EBERSOLE: It is not hypothetical. I just l 12 want to know.
13 Would you say that the gaality of the arguments 14 which are used to defend the thesis that those pipe breaks 15 wouldn' t occur was simply lower than the quality of your j 16 arguments?
17 MR. O'BRIEN: Most piping that we have had 18 pressure boundary failures, I think we found the plants were t
19 not qualified for our acceptance criteria. Yes, that is for i 20 sure.
21 We didn't ignore the operating history.
22 (Slide.)
23 Finally, the Japanese are very assertive in going 1 24 forward with leak before break, and the Canadians are also 25 applying it at Darlington. Most other countries are simply I
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1780 03 07 34
) DAVbur 1 looking at the others.
2 DR. SHEWMON: When you make that statement about 3 the Japanese, does this mean they are taking out pipe whip 4 restraints, or what evidence of assertiveness do you see?
I 5 MR. O'BRIEN: They are applying it to boilers.
6 That is something that we would probably have a little 7 difficulty to do right now.
8 I don' t know, they have not done it yet. You 9 see, the Germans have done it, okay. They have units 10 operating without pipe whip restraints in locations where we i
- 11 have pipe whip restraints.
! 12 MR. MICHELSON: Is this outside of containment?
l s-} 13 MR. O'BRIEN: No. The answer to that question is 14 probably hard to read.
l 15 (Slide.)
16 But you have a copy of it.
i 17 MR. MICHELSON: It just says FRG on there.
18 MR. BOSTAK: The information we got from them at i 19 the meeting was that the Japanese were undecided with f 20 respect to older plants.
21 We asked each country's representative there, and 4 22 I think that may have been furnished to you in one of the i
1 23 reports that Mr. Rodebaugh furnished, but we asked each one l
, 24 what their intent was with respect to older plants.
< (
25 MR. MICHELSON: My question was outside i
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1780 03 08 35 DAVbur 1 containment, not older plants.
2 MR. BOSTAK: As far as outside of containment, we 3 didn' t limit it. In other words, when we got to other 4 piping -- and that included outside containment -- that was 5 a question, and we did receive answers on other piping 6 systems. Some were new and some were old, but the Japanese 7 had a conditional yes.
8 MR. MICHELSON: Outside of containment?
9 Thank you.
10 MR. REED: I would like to point out that that 11 upper righthand block, if parts of it were extracted and put 12 up front as what the proposed rule is, I don' t think we 13 would have had all the argument that we have had because if 14 that was appropriately written at the outset of the
- 15 paragraph, the proposed rule, we would see that BWRs are not 16 included -- and I am happy about that -- and that piping 17 outside containment is not included. And I am happy about 18 that because there is no real leak monitoring as there is i
19 inside containment.
l 1 20 MR. O'BRIEN: That would be totally wrong though, 21 but we intend to apply this to boilers if they meet our 22 acceptance criteria.
23 MR. REED: But you say only the BWRs meet it 4
24 right in the bottom sentence there.
(:) 25 MR. O'BRIEN: Now at this time. We have not 1
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1780 03 09 36 DAVbur 1 evaluated other piping.
2 MR. REED: If you do evaluate and you come out 3 with that boilers meet it, then I will be very unhappy.
4 MR. MICHELSON: Broad scope means everywhere, 5 inside and outside.
i 6 DR. OKRENT: Well, I said a pig in a poke.
7 MR. REED: Well, I will have some additional -
2 8 comments on this one.
9 MR. O'BRIEN: You see, you are concerned about 10 our acceptance criteria. Right now they would not need it, 11 but suppose they changed the piping , right?
12 MR. REED: I don't think that changing the piping 13 on a BWR is going to change the oxygen environment, and I 14 don' t think until you have got 10, 15, or 20 years that you 15 have any idea wherbsc it is going to meet those cracking 16 situations.
17 DR. SHEWMON: That is one place where the Germans 18 and the Japanese are doing things that we would not.
19 Okay, go on.
20 MR. O'BRIEN: On reporting foreign practices, it 21 changes, as does ours. So this is a few months ago, and we
! 22 just learned a little while ago, the Germans are now 23 applying leak before break to the boilers, and they are 24 applying it to older reactors. But as of a few months ago O 25 titey had limited it to PWR, primary loop and pressurized ACE-FEDERAL REPORTERS, INC.
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'(_) DAVbur 1 by main steam, main feed, some RHR piping, only inside the 2 containment and where we allow to contain the piping 3 -anywhere.
4 MR. REED: I think that is very sensible policy, 5 state of the art. The German policy.
6 MR. O'BRIEN: But that has changed already. This 7 was a few months ago. Now we have learned just recently 8 that they are applying it to boilers.
9 DR. OKRENT: Again, you make a statement, they 10 are applying it to boilers, but you are not specific, and 11 without saying in exactly what way and under what 12 conditions, with what requirements they are applying it to 13 boilers, to me it is --
14 DR. SHEWMON: A pig in a poke, to coin a phrase.
15 (Laughter.)
16 DR. OKRENT: In fact -- well, I have a stronger 17 word.
18 MR. O'BRIEN: Anyway, we apply any piping to any 19 reactor type.
20 DR. OKRENT: Excuse me. I want to now get to the 21 water hammer question.
22 How do you decide whether a steam line will have 23 water hammer in it?
24 MR. O'BRIEN: Operating history. There are two O 25 ways we allow them to address water hammer problems. One ACE FEDERAL REPORTERS, INC.
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(_/ DAvbur li j is they document that specific line that has a good history 2 of not having water hammer or else take measures to mitigate 3 - or prevent water hammer.
4 MR. REED: But if steam generator overfill of a 5 EPWR is possible, then you must admit that water hammer is s
6 possible.
, 7 MR. O'BRIEN: They probably won' t make it if they 8 don't. We have this extremely low probability criterion
, 9 that they have to satisfy for water hammer, corrosion, 10 creep. We have a big list, by the way, 11 MR. MICHELSON: We used to have a concept called 12 defense in-depth. What we are doing is taking some of the 13 depth out of defense.
14 MR. O'BRIEN: We challenge that.
15 DR. OKRENT: I don' t want to leave the water 16 hammer question.
.n 17 MR. O'BRIEN: How can you say that we have never s
18 needed this' with such a long history?
19 MR. MICHELSON: I was just citing the example.
20 DR. SHEWMON: Let's go back to Dave. He had the r
21': floor and the question.
22 DR. OKRENT: Until let's say the fall of '85, San 23 Onofre had not had a severe water hammer. Okay?
24 So you would look there, I assume, through them?
O 25 MR. O'BRIEN: No, we wouldn' t because there were ACE-FEDERAL REPORTERS, INC.
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( DAVbur 1 a lot of previous water hammers at Indian Point and 2 elsewhere and at Maine Yankee. There is a feedwater line.
3 We have a high expectation of water hammer in feedwater 4 lines.
5 DR. OKkENT: So you are saying these main and aux 6 feedwater lines are excluded for all reactors? ,
7 MR. O'BRIEN: Unless you have done something like 8 putting damper-valves. If he takes measures, steps to 9 mitigate water hammer and the staf f evaluates those measures 10 as adequate enough to reduce the probability of water 11 hammer, as it stands right now feedwater wouldn' t make it.
12 MR. MICHELSON: I remind you that the check 13 valves should have done the job -- at San Onofre they 14 f ailed -- because they had failed much earlier in the time 15 history, and we never predicted those kinds of multiple 16 failures, and so you didn't predict the blowout.
17 MR. O'BRIEN: But we didn' t need pipe whip 18 restraints at San Onofre or anywhere. We didn' t get a pipe 19 rupture there.
20 DR. OKRENT: Now, see, you are relating it all to a
21 pipe whip restraints, though. There are other issues that 22 could arise as well.
23 Again,-the routing of lines -- maybe it is not 24 near the control room, but it is near a lot of other control 7,,
V 25 centers.
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s,_ / DAVbur 1 I don't know why. Okay?
2 You have remcVed the need for the design or to 3 think about it. You have removed the need for the staff to 4 review it.
5 DR. SHEWMON: It is all under the dynamic 6 effects.
7 MR. MICHELSON: Including pressure, though.
8 DR. SHEWMON: No, pressure transients, not 9 pressure. They have told us that the pressures are still 10 there.
i 11 Okay, Mr. Kerr.
12 DR. KERR: My impression, having listened to this 13' in subcommittee, is that the staff believes they will make 14 things safer. I haven't heard any comment on this.
15 Carl and Dave may disagree that it is possible to 16 enhance safety by the approach, but the impression I get is 17 that this is felt to be something that only relaxes the 18 requirements and therefore can' t have a positive ef fect and 19 can only have a negative effect.
20 DR. OKRENT: I don' t think you understand my 21 position.
22 DR. KERR: I am not claiming to understand your 23 position. I am trying to interpret what you said.
! 24 DR. OKRENT: I am trying to look at what the 25 Germans did. I am trying to figure out why did they do it, ACE-FEDERAL REPORTERS, INC.
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1780 03 14 41 U DAVbur 1 what did they do, and I can see a lot there that in fact 2 sounds like a good approach.
3 DR. KERR: But you don't see any possible 4 improvement in safety from this approach. What bothers me 5 is I don' t see the staf f trying to accomplish, le t's say, in 6 an adequate way what the Germans have tried to do when they 7 did remove the restraints. They are talking about 8 restrictive criteria, and so forth. I don' t know what they 9 are.
10 I don' t know whether in f act it is possible to 11 have adequate inspection or leak detection, and so forth and 12 so on.
b) t/ 13 So I earlier said I might well go for some 14 ' package that removed restraints, but I don't right now know 15 , what they are going to allow, and I have seen enough of the k
16 workings of the staf f over 20-plus years to fear a movement l
17 which is not in the general direction of safety.
18 ,
19 20 21 22 23 O
'v' 25 ACE-FEDERAL REPORTERS, INC.
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1780 04 01 42 f-(_y) DAVbur 1 MR. MICHELSON: Since my name was used in vain, I 2 would like to clarify, I do not in any way object to the 3 movement of restraints. I am all for it. That is not my '
4 problem at all.
5 My concern was only with the design of the ;
6 subcompartments outside of containment, which I view much !
1 4
7 like the containment itself. It helps to direct the steam l l
8 and everything away from vital equipment out of the 9 building, and I don' t want those walls to blow down or to be !
10 violated in the future so that there are now holes in them 11 where there used to be nice, big, solid compartments.
12 DR. SHEWMON: I think the staff criteria are 13 fluid enough so that they are interested in our comments.
14 Try to get on through this, if you would, John.
15 MR. O'BRIEN: A couple of months ago the Germans 16 were only applying it to new piping. Now they are not.
17 Age is not a factor in the United States. They i 18 did not apply it to containment, EQ, or ECCS. Neither did 19 we.
20 They retained dynamic loads on components. Our 21 decision is to remove those forces and certain of our 22 licensees are already taking advantage of this. They are 23 removing snubbers from supports. It is a very big 24 decision.
-O 25 That one right there. There is a distinct i
ACE-FEDERAL REPORTERS, INC.
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._ DAVbur 1 difference in the way they are doing business. They have a 2 10 percent flow area replacement crack, but the change in 3 their position is that they allow fracture mechanics to 4 assume something other than a 10 percent flow area.
5 We have nothing except that we still have breaks 6 in connecting branch pipes. This, I think, is what 7 Mr. Michelson was pointing out.
8 They retained the dynamic effects of 9 pressurization in compartments, in cavities. We are 10 eliminating them except in those volumes that are part of 11 the contaiment function.
~
12 They have something very curious on decompression
_J waves within ruptured pipes. They eliminate it on the 13 (
14 primary side, but they Keep it on the secondary side. We 15 think the reason for that is that want some design basis for l
16 P the steam generator. We eliminated them if we can show a 17 leak before break on both sides.
18 Finally, they have even more stringent 19 requirements on leak detection than we do. They have five 20 ways of detecting leakage. We have three.
21 That is already out there in summary of the i
22 German versus American practices, or proposed American 23 practices.
,3 24 (Slide.)
}
~~'
25 My next slide deals with value impacts. They are ACE-FEDERAL REPORTERS, INC.
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quite large.
2 The last bullet, the one that is telling, we are 3 talking about $100 million per unit and the man-rem are 4 quite large. They run to up to 300 to 800 man-rem per 5 plant. That is over the entire life of the plant.
6 And in addition to what is being accomplished, we 7 are taking major strides forward in eliminating pipe whip l
8 restraints and dynamic effects as we offer provisions to 362 9 dealing with breaks. It turns out that the value impacts 10 associated with that -- which is going forward already 11 without your review --
12 (Slide.)
)0 13 --
there may be about 20 or 30 plants that have 14 already taken advantage of it.
15 . And then my next slide deals with NRC resource 16 demands, and we need 10 to 20 man-years to deal with this.
17 And we promised to do this in the limited scope rules. The i
18 Piping Committee advocated this, and we have got a couple of 19 applicants in already.
20 MR. REED: I might point out, I believe the ACRS 21 GESSAR letter says that you can't use the leak before break 22 rule for GESSAR, is that correct?
23 I think so. You should check the ACRS letter.
,- 24 DR. OKRENT: Can' t. All we can do is recommend V 25 anyway. <-
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(_/ DAvbur 1 DR. SHEWMON: I don' t remember that. Can we get 2 a copy of the GESSAR letter?
3 DR. OKRENT: GDC will, anyway, I think guide the 4 staff.
1 5 MR. MICHELSON: Have you given any permission to l 6 use the broad scope rule? l 7 I guess you haven' t. It is not a rule yet.
8 Nobody is changing outside of containment yet.
9 MR. O'BRIEN: No. We have given exemptions to 10 those limited scope rules to a few plants, but nobody has 11 taken advantage, although one major utility is going to be 12 in a real pickle --
(O V 13 (Slide.)
14 --
if they don' t GDC for broad scope because they 15 are going to bear huge expenses if Beatrer Valley doesn' t get 16 it.
17 MR. MICHELSON: Is that relative to supports, 18 pipe whip restraints?
19 MR. O'BRIEN: If they are building the plants 20 without many, it is just a small fraction.
21 I spoke very briefly about our work on revising 22 362 again with arbitrary leak before break, and it is just 23 peripheral to this action.
24 (Slide.) l
[
~ i 25 I think that is all I wanted. I didn' t bring my l
l ACE-FEDEftAL REPORTERS, INC. l 202-347-1700 Nationwide Coserage 8aL336-6646 1
4 1780 04 05. 46
(~\
s_) DAVbur 1 slides on agceptance criteria because I thought my time was i
2 limited. But I would be pleased to talk to you about it 3 because that is really the central issue.
4 It will give you all a warm feeling because you 5 are going to see how difficult it is to take advantage of 6 leak before break technology. 'I think the fears that I see 7 here in this group are mostly resulting from the fact that
- 8 you think we are going to capriciously implement the rule.
9 It is not that way at all. There are several 10 hurdles you have got to jump before you can apply the rule, 11 and a lot of piping is not going to make it.
12 MR. REED: I think there are two issues here 13 affecting boilers -- piping and the high rate cooling 14 environment in oxygen.
15 DR. SHEWMON: Let me go back on this. I haven' t
- 16 heard the staf f say anything about approving GESSARs as 17 built now.
18 Is that right?
19 But if under the revised 0313, which we will 20 discuss in an hour, they might allow if they could be i 21 convinced that they had at least two ways, in effect, for 22 reducing or eliminating the probability of stress corrosion 23 cracking.
24 One of these possibilities would be hydrogen 25 treatment. The other would be replacement of lines with ACE-FEDERAL REPORTERS, INC.
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)~.DAVbur 1 a stress corrosion cracking resistant material, but that 2 . still is scuething the staf f has not signed of f on yet, and 3 it is only a matter of individual comments on what they 4 think might be the criteria.
1- 5 So for now the position is that BWRs are excluded 6 from this.
5 7 MR. O'BRIEN: They do not satisfy acceptance n
8 criteria.
i
- 9 -
MR. REED: Rigorous acceptance criteria, yes, I 10 can buy that, and I hope that upper block that you had in i
11 the upper righthand corner, that is fine and dandy but I am 12 .a little worried about who is going to be writing that
- -\ , 13 criterion and implementing it.
! 14 Therefore, I wish a broad statement up front.
15 MR. O'BRIEN: If you have a copy of our
- 16 acceptance criteria -- aro in our package -- we published it 17 for comment. There is one section of the Federal Register 18 notice that is also published by the Piping Review Committee i
19 in NUREG-1061, or Volume 3.
20 MR. REED: You are saying that all piping is okay 21 subject to our review by rigorous acceptance criteria.
22 Why don' t we say that certain piping is okay 23 because it has proven itself.in 40 years in the field and ~
24 all these kind of things and some of the piping is not okay 25 until it has proven itself? -
4 4 ACE-FEDERAL REPORTERS, INC.
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1 1780 04 07 48 DAVbur 1 y MR. O'BRIEN: I think even primary coolant is 2 okay until you prove it. It is very restrictive.
3 I am asking you to look at it. I am sorry that 4 the full committee didn' t get this package until right i
5 today, although we have published it for more than a year, 6 our acceptance criteria, and we have spent a lot of time 7 debating it and making sure that there were no loopholes on 8 that.
9 It is going to be hard to get over those i 10 acceptance criteria. One section of the rule is the 11 acceptance criteria. I don' t think anybody is going to feel 1 12 real comfortable until you look at it and see how tough we 13 are. I promise you we are tough.
t 14 MR. WARD: Dr. Shewmon, maybe the members could 15 read that over lunch.
16 DR. SHEWMON: You said I had three hours. I will 4 17 stay with it.
18 MR. WARD: That is not all we said.
19 MR. MICHELSON: Unless you can apply leak before 20 break -- that is the part I am interested in -- even for 21 piping, abnormal pipes, therefore you can essentially show 22 leak before break, there will be no leaks, no breaks.
4
- 23 MR. O'BRIEN: But you have to demonstrate that.
l 24 MR. MICHELSON: I understand. I am just trying C:) 25 to understand, though, when you first started out with ACE-FEDERAL REPORTERS, INC.
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-1780 04 08 49 DAVbur 1 the intermediate,- now we are actually talking .about the 2 terminal anchor points as well where the stress situation is 4
3 somewhat dif ferent and generally a lot higher, and many 4 interesting problems.
- 5 DR. SHEWMON
- If the stress is a lot higher under l
6 the criteria they are suggesting, you are perfectly free to 7 read it. I would encourage everybody to do so. We will 8 discuss that. ,
9 Right now the staff is talking about extremely
) 10 low usage levels.
i 11 MR. MICHELSON: I just wanted a clarification 12 about the fact that there aren't any breaks, even at k 13 terminal points that show leak before break.
I 14 The next clarification has to do with the fact 15 that piping is generally classified as pipes as opposed to 1
16 bellows as opposed to other kinds of components.
17 What is the staf f's position about treating these 18 other components? Are there any breaks anywhere any more, 19 even on a bellows for instance?
- 20 And by the way, there are many high pressure t 21 bellows, keeping in mind there are a number of high pressure l 22 water systems over 275 in nuclear power plants. Go and I
23 look. They aren' t all steam.
t ,
24 MR. BOSTAK Again, we are talking about what we C:) 25
, have today, but at the current break locations they are f
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~(_) DAVbur 1 located in pipe. They are not located in these other 2 fittings, and they are not located in bellows.
3 MR. MICHELSON: That is my understanding.
4 Now, you are still saying that even with leak ,
5 before break you are only dealing with pipe.
6 MR. BOSTAK: We are not coming up with anything
- 7 new. We are not replacing this with breaks someplace else.
I 8 MR. MICHELSON: When you eliminate, though, the i
- 9 pressurization of the compartment, which is fine --
10 MR. BOSTAK: The dynamic effects only. If you 11 have leakage -- and you do. You always have. t i 12 MR. MICHELSON: My concern is that we used to 13 say, okay, you know, if you retain the pressure design 14 characteristics, it would take care of this bellows question 15 where the bellows is in the compartment.
l 16 If you now eliminate it, because you have 17 eliminated the break on the pipe, you haven' t taken care of 18 it.
19 MR. BOSTAK: We are not saying now and we didn' t 20 say before that bellows are going to rupture. You may have
! 21 leakage, and you will have leakage there.
22 MR. MICHELSON: Well, you know bellows as well as i
23 I do. They do rupture, and they rupture catastrophically.
I 24 They don' t follow leak before break theories.
4
(:) 25 MR. BOSTAK I have seen f atigue cracks, but I i
i ACE-FEDERAL REPORTERS, INC.
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1780 04 10 51 2 DAVbur 1 haven' t seen any that catastrophically have ruptured.
2 MR. MICHELSON: That is interesting.
3 DR. SHEWMON: Are there any other questions at 4 this time?
5 (No response.)
6 DR. SHEWMON: Okay.
7 If I hear no other questions, I suggest we take 8 an hour break for lunch and come back at 10 to 2:00.
9 (Whereupon, at 12:50 p.m., the committee was 10 recessed, to reconvene at 1:50 p.m., this same day.)
11 12 c3 ;
w- 13 !
i 14 l l ll 15 l 16 i b i 17 l l l
18 '
1 19 20 l
21 22 l 23 l l
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- z. ( j AFTERNOON SESSION 2
(1:50 p.m.)
3 MR. WARD: Dr. Shewmon. p 4
MR. SHENMON: Let me read two paragraphs out of
' the Summary of Acceptance Criteria that I should#have had 5
6 in my introduction.
7 "While the NRC acceptance criteria are subject 8
to further. elaboration and revision as the results of 9
ongoing studies become available, some details on existing je staff guidance, based primarily on Chapter 5 of Volume III i i n of NUREG 1061 -- this was the pipe study group chaired by i
! 12 Spence Bush that was put out in November '84 -- some l i3 details are given below."
14 The first paragraph says "The leak before break 15 approach should not be considered applicable to fluid 16 system piping or portions thereof that operating 17 experience has indicated is particular susceptible to ,
18 failure from effects of corrosion, water hammer -- in
! 19 fact, they call out specifically intergranular stress i
20 ' corrosion cracking -- water hammer, creep or low or high
- 21 cycle fatigue, i.e. thermal mechanics.
22 "To show that piping systems are not susceptible 23 to failure from water hammer, creep and corrosion means a
24 extremely low probability criteria must be satisfied."
25 This is a low probability criteria for failure i
53 t\
(_)'
I which is mentioned explicitly in the rule.
2 "This can be accomplished through investigations 3 of operating history and measures to prevent or mitigate
.4 these phenomena. Creep is not a problem in light-water 5 reactors, but may occur in other reactor types covered by 6 this rule."
7 MR. EBERSOLE: Did he have in there inadvertent 8 overpressurization?
9 MR. SHEWMON: Not explicitly, though if there 10 was any history of this, it would come in under the I
11 operating experience that they are talking about.
12 MR. MICHELSON: And, unfortunately, you may only i
- get one event in the operating history before this all 13 14 gets very interesting, because we don't postulate check ,
15 valve failures when we determine whether it is acceptable 16 to water hammer and things of that sort.
17 MR. SHEWMON: Let me bring up one other thing 18 that Ed Rodabaugh, who has been an expert on the piping 19 code for a long time said he had never seen a guillotine 20 break in stainless steel, and out of thousands one or two 21 in ferritic, i
22 So I think the chances of your getting a sudden 23 break, even if you did overpressurize it, of the scale is 24 a very low probability, like vanishingly small. What the ,
- 25 staff does about indeed coping with leaks or sprays or
54
_; I jets from fatigue cracks is something I think they should 2 look at hard and I think they will.
3 Are there other questions on this?
4 MR. OKRENT: I will just offer the comment. I 5 did read it during lunch.
6 MR. SHEWMON: The criteria. i 1
7 MR. OKRENT: Yes, and I find that they are not 8 really what I would call stringent, and I think they can 9 be used by the staff to be stringent or non-stringent as 10 they are written. I will put it that way.
11 I think the strong emphasis on operational 12 events leaves me feeling ---
,- 13 MR. SHEWMON: Uneasy.
1 I" 14 MR. OKRENT: Uneasy. That is exactly the right is word. And I don't see the compensating things that the 16 Germans have built in, but they built in something to take 17 away that uneasiness. This is what I am getting at, and 18 they have a much more limited, at least with what they 19 have so far permitted, a much more limited applicability.
20 The staff seems to be jumping in a general way, 21 and it is hard to say just what the specific 22 implementation is. Just like with the safety goals we 23 were talking about, there are some general goals which we 24 will probably all be able to approve, but it is that 25 matrix that in the end is going to be the thing that the 8
i I
55
.rs.
(_) i staff uses.
2 MR. SHEWMON: Well, there is a letter which has t
3 been handed out ---
4 MR. IGNE: Not yet.
5 MR. SHEWMON: Would you hand it out. --- which
, 6 you can read between now and tomorrow morning. As you can
, 7 see there, my concern is that the staff is going to 4 8 require so much analysis and paper and fussing that it 9 would be easier to leave the crazy things in. So I am 2
to concerned on the other side of your concern.
11 MR. OKRENT: But, again, I am not arguing in 12 favor of leaving something in. I am trying to find out 13 whether we will have an acceptable situation, or whether 5J 14 we in fact do have an acceptable situation.
15 You know, if they are not doing any good where T
16 they are, there may still be situations that the staff
- 17 isn't dealing with adequately, whether it be by leak
- 18 detection where they are not going it or whether it be by 19 thinking more about where water hammer might occur and l 20 seeing what steps can be taken to anticipate instead of 4
21 relying on it. I don't know what it is. Uneasy is the 22 right word.
23 MR. SHEWMON: -Okay.
! 24 MR. REED: Well, I only saw the summary of
- 25 acceptance criteria. Some of the criteria looked
56 L j 1 rigorous, and then I have trouble with this one. "The 2 leak before break approach should not be considered 3 applicable to fluid system piping or portions thereof that 4 operating experience has indicated is particularly 5 susceptible to failure from effects of . . . . I don't 6 like "particularly." I would like to say "even slightly."
7 MR. OKRENT: That is not what I call stringent.
8 MR. REED: I don't call it stringent either.
9 MR. SHEWMON: Well, since there haven't been any 10 failures of this sort, I assume they are talking about 11 susceptible to leaks.
12 MR. REED: Just say intergranular stress g s. 13 corrosion cracking. It has got to be particularly i l
'd 14 susceptible.
15 MR. MICHELSON: There have been some rather 16 large ruptures of relatively high pressure water lines 17 that have dumped a hell of a lot of water in a hurry, and 18 that is the kind of rupture I am more concerned about than 19 the big, double-ended rupture, which I think is extremely 20 unliP.ely. But these pressurizations of compartments, that 21 being of water as the fluid, as soon as that compartment 22 is full, that fluid system pressure, whatever it remains 23 of the system pressure that is pumping on the pipe, is now 24 applied to the building and it pulls it apart. It doesn't 25 take many poundo. A few pounds is all it takes to blow a 8
57 I compartment out.
2 MR. SHEWMON: I agree.
l 3 MR. MICHELSON: So I am kind of concerned about.
! 4 our design basis to make sure.there are enough drain. lines ,
~
5 and blow-out panels or something-to take care of these 6 things.
7 MR. SHEWMON: Nobody is talking about taking out 1
i a drain lines, but you can get back to your question of how j 9 do you know-they aren't going to change things in the 1 to future.
! 11 MR. MICHELSON: How do you know they aren't l
i 12 ' going to close the drain now that this is over with. We 13 have got this huge blow-out panel and it is giving us some i 14 kind of other difficulties for some reason and we take it i
j 15 out. It is not required.
i i
16 MR. SHEWMON: Does anybody from the staff have 17 any closing comments?
18 MR. BOSNAK: Well, I think all of the doctrines 19 that you are looking for will be developed by the staff.
20 We are waiting to get public comment on this. We have 2 called out in particular several areas that we are looking 22 for comment on.
23 We are not looking for people to change things, 24 as Mr. Michelson mentioned. I know that is in the back of 25 your minds that you are afraid that might happen, but you
- lif
- - 58 1 have to rely'on some degree of engineering judgment being 2 used both by the people out in the industry and by the 3 staff, and eventually we will have such a document that 4 you will be able to review in detail.
5 MR. REED: Coming from the operating plants, I 6 realized tremendous problems associated with restraints
! 7 and bash pots and their testing and tremendous problems in
< 8 radiation exposure and all these things.
9 I have the feeling that'any licensee who sees l
10 somebody waving just a bit of the opportunity to get-rid II of some of these restraints and bash pots will seize upon 12 it, and I have feeling that if you go out with this to the 13 public, if there are weaknesses in it and not rigorous 2
-"d l 14 aspects in it, you are going to deluged with let me do i
15 this tomorrow.
16 MR. BOSNAK: Well, we have already started this, t j 17 as we mentioned, with respect to Beaver Valley, and we i 18 have had many meetings.
l 19 And I think, as Dr. Shewmon has mentioned, he
- 20 probably has heard some of the comments perhaps from the 21 applicant that the staff is really going over on the other 22 side of the line to make sure that they are not applying ;
23 these things where they should, i 24 Right now we are feeling our way.
25 MR. REED: But don't you have to write this in 111 n .,,---,--r-,- ,,.-rw-.--n,,.,n,., ,.rene---
e n c .c , s ---,.,,~,,,,wy-. w.,..en. e,e..,.<e-,,,- , , + - ,.e,ne, - ee,<.,
l 59
, 1 such a way that it is attractive looking at the outset?
2 MR. SHEWMON: They are changing two sentences, 3 and if I have my way, it will only be one sentence on the 4 end of GDC-4. So that is a long ways from the other end 5 of the piece of spaghetti. I don't know what the statuo 6 of these 20 pages of comments that come with it are. So, 7 you know, how it gets implemented is something that we 8 have an interest in and we will pursue and should.
9 MR. BOSNAK: We would like a letter to have this 10 go out for public comment. That obviously the staff would 11 like very much.
12 MR. SHENMON: Okay.
c~ i3 MR. HERNAN: Mr. Chairman?
( /)
14 MR. SHEWMON: yes.
15 MR. HERNAN: One real quick point. Mr. Reed has 16 a problem with the words "particularly susceptible." I 17 would like to point out that Dr. Kasner, who is one of 18 your subcommittee consultants, made a big issue of the 19 point that no material is really totally susceptible to 20 corrosion, and it is a degree of how susceptible. Some 21 materials are more susceptible than others.
22 MR. SHEWMON: What was the word that came after 23 "Particularly."
24 MR. REED: Well, to failure from the effects of 25 corrosion, such as, intergranular stress corrosion.
8
60 f~l
(_J 1 MR. SHEWMON: What was the word that came after 2 "particularly susceptible"?
3 MR. REED: "To failure."
4 MR. SHEWMON: Okay. It is a long ways between-5 having cracks being observed in the material and having 6 failure. So I guess we can sit here and say we are 7 against things that are particularly susceptible to 8 failure, and that isn't separate from the question of 9 whether or not anybody has ever been able to crack the 10 material in some set of lab conditions. We will find lots' 11 of things to argue about before we finish with this.
12 Gentlemen, I would like to move on. There is a 13 letter here that I have proposed to the committee. As Dr.
14 Kerr likes to say, it is now the committee's letter. I 15 have asked the Chairman to take it up in the morning. You 16 can read it and see whether you would like to suggest 17 something else or if you think that would do it.
18 I would like to move on at this point to 19 something which I think will be easier and less '
20 controversial, and this has to do with another by-product 21 of the piping study group having to do with stress 22 corrosion cracking and how the Commission copes with this.
t 23 There is a reg. guide 0313, which has been out 24 once and has been revised once and now has been revised 25 again. This has been available for study and comments for 8
I l
l 1
i
61 1 the last several months.
2 What I am having handed out are tuo pages from 3 the staff presentation tha* we heard at the subcommittee 4 meeting which I think summarizes what the changes are, and 5 I would like to go over that and, if there are questions, 6 we can ask them of the staff. There is a staff 7 representative here.
8 The committee's position on this was that we 9 would urge its approval. The letter that you will see in 10 a little bit has one significant comment in it, and that 11 is that there should be a limit on the number of welds in 12 the system that can be cracked, whether they are repaired 7s 13 or not. But let me go over what we have got here.
! l la The Rev. 2 expands Revision 1 coverage to all 15 stainless steel piping, Classes 1, 2 and 3. Rev. 1 had 16 been limited I think probably to Class 1 or to only the 17 primary system.
18 The second revision will require formal 19 qualification of NDE examiners and procedures, and this 20 will be done at the EPRI NDE conter. Rev. 1 recommended 21 that improved techniques be used that didn't require the 22 formal qualification.
23 MR. WYLIE: you say at the EPRI NDE center. Is 24 that the only place they can be qualified, or is that just 25 a place?
8
62
. ,j 1 MR. SHEWMON: It is my impression that is the 2 only place they can be qualified. It is not something 3 where everybody has their own failed things and can check 4 out their inspectors on them there, which is sometimes 5 used in other situations.
6 provide guidelines for evaluation and repair of 7 cracked welds. Rev. I required replacement of cracked 8 welds. There is now a code approved case, I guess it is, 9 though I am not positive that is the correct nomenclature, 10 on how to evaluate and repair these cracked welds.
11 As I say in the letter. we have suggested there 12 should be a limit on this, but otherwise going along with
, s, 13 it, i ,
\ ]
'" 14 On the second page there is a list of 15 recommendations and then that is a list of four options 16 that can be used, and the staff then will allow release of 17 inspection if these things are implemented. They feel 18 that they wish to encourage them to put in material 19 resistance to cracking and that the stress improvement 20 process will give it the stress corrosion cracking and 21 that the water chemistry improvement will. So they will 22 give them credit for all of these as well as repairs. I 23 guess I am less clear on that, what they buy there.
24 There are guidelines for crack evaluation and 25 repairs, upgrades, leakage limits and monitoring. In t
63 1 general we thought it was a good package and would approve l 2 --
it is approved going out for comments, or this is not 3 for comments but is the issuance of a ---
4 MR. IGNE: This is going to go out for comments.
5 MR. SHEWMON: Okay. So again, it is for 6 comments.
7 Any questions?
e (No response.)
9 MR. MARK: Out of ignorance, Paul, I am sorry, to but the inspection, crack assessment has I believe 11 improved over the last three or four years considerably, 12 and did you say the EPRI qualification for inspectors is to be invoked here?
9 13 14 MR. SHEWMON: Yes, as part of their center. The 15 staff certainly works with them on what are acceptable 16 criteria.
17 MR. MARK: Yes. And they will be checked out la for exercising the best procedures that people now have in 19 hand?
20 MR. SHENMON: Not only exercising them, but 21 exercising them effectively. I think the bottom line is 22 can they find flaws when they are there, do they overcall 23 and can they evaluate them on size.
24 MR. MARK: All right. So in the pattern as it 25 is being set up, that kind of thing will be ascertained?
8
64 7-.
t_ , 1 MR. SHERMON: Yes, and rather stringently in 2 fact because a lot of the industry inspectors were not 3 passing for a while, and I don't know how they are doing 4 now.
5 MR. MARK: Yes. Well, that is the vague 6 recollection I had. Now then there are places where 7 applying these procedures are particularly awkward. What 8 is said about that, you know, angled joints and places 9 where things have been butted together?
10 MR. SHERMON: There are places where if they 11 can't get at it, they can ask for a waiver.
12 MR. MARK: Well, either ask for a waiver or come
,s 13 out not really knowing the state of things that are there.
! )
14 MR. SHERMON: Would you comment on that 15 question?
16 MR. KOO: That is for those welds not accessible 17 to UT, then they will have to perform by UT.
18 MR. MARK: So what should I think? On some of 19 these you will really not be able to say there are no 20 cracks here bigger than thus much.
21 MR. KOO: Yes, there are some welds in the wall 22 that, right, cannot be inspected.
I 23 MR. MARK: Is there a change in the way on l 24 approaches those situations implied there? i 1
25 MR. KOO: Those welds can be UT inspected, yes. )
lli l
65 O
(_) 1 MR. SHEWMON: His question was has anything 2 changed between Rev. 1 and Rev. 2 for these inaccessible 3 welds?
4 MR. KOO: No change.
5 MR. MARK: What do we do about them as it 6 stands? Do we say well, we can't see it, so we hope it is 7 all right, or hit it with a hammer and see if it sounds a 8 little cracked or what?
i 9 MR. SHEWMON: You practice leak before break I l
to think because that is the only defense you have at this l 11 point. Is that correct?
12 MR. KOO: Yes.
-m 13 MR. MARK: And that remains as it has been.
14 MR. KOO: Yes.
15 MR. MARK: Okay. Thank you.
16 MR. SHEWMON: Other questions?
17 (No response.)
18 The third thing that we had then at the i 19 subcommittee meeting -- and if somebody would try to find 20 Professor Okrent, I think it would be appropriate at this 21 point in time -- we went over the questions that had been 22 in this draft letter on PTS, and in the minutes, or what 23 is in your folder under Tab 10 is a writeup of what Igne 24 took as the questions and answers that were in the 25 letter. This is on page 13.
66
_j _
1 I don't know quite how to handle these, but I 2 think they cope with them reasonably well, but then I 3 wasn't the one to be convinced of this earlier.
4 MR. REMICK: Dr. Shewmon, while we are waiting, 5 does the committee plan to issue a separate letter on Reg.
6 0313 Rev. 2 for comment?
7 MR. SHEWMON: Yes. I do not have a draft of a 8 letter on PTS. (a) I want to write a letter and (b) I see 9 absolutely no point in combining it with anything else.
10 (Pause while waiting for Dr. Okrent to return.)
11 MR. REED: I would like an explanation for A-1.
12 MR. WARD: Do you want the explanation for A-17 g, 13 MR. REED: Yes, I will need an explanation.
' 14 MR. WARD: Well, it is granted that the 15 challenge rate -- are you talking about on page 137 16 MR. REED: Yes.
17 MR. WARD: As I understood it, let me just put 18 it briefly, that is is granted that the challenge rate 1 1
19 would be higher, but in the B&W plants back flow through 20 the reactor vessel vent valves tends in most scenarios to l
21 provide more of a mixed fluid. l l
22 MR. REED: Not unless the plugs are stopped. l 23 MR. WARD: Yes. Well, that is in most scenarios l 24 where you are worried about pressurized thermal shock.
25 MR. REED: It is?
8
67 1 MR. WARD: yes, and the judgment is that it sort 2 of tends to balance things out, and that they think 3 overall that the problem isn't any worse in B&W plants 4 than in others.
5 MR. SHEWMON: Dave, on page 13 under Tab 10 is a 6 list of the questions and a summary of the answers.
7 MR. OKRENT: I read that. Where are we, I am 8 sorry.
9 MR. SHEWMON: We are on question 1 now, and 10 Glenn wanted an explanation of the answer to question 1.
11 Why don't I ask, Roy, if you would go over this. The 12 question was "Is there reason to believe the issue of PTS
,- 13 cannot be treated generically?" Some classes of plants of i !
14 particular designs are subject to a significantly higher 15 frequency of overcooling events." Can you give us an 16 answer or a comment on that again?
17 MR. WOODS: This is the basic question the la Commission has asked us when we first proposed to have a 19 PTS rule back in 1982. And as a result of that question, 20 which was should we or should we not apply the same 21 screening limit to B&W plants as we apply to everyone 22 else, there was a B&W owners group task force formed.
23 They did a PRA analysis of a " generic" B&W 24 Plant. Basically they found that although the challenge 25 rate might be somewhat higher, in other words, the 8
68
. _j 1 frequency of overcooling events, mild overcooling events 2 might be higher, but for a lot of the significant events, 3 not all of them, but a lot of them, the result when you 4 lose force of circulation, the RCPs, and on B&W plants for 5 that particular class of events you do have the vent 6 valves which provide significant mixing and thereby 7 prevent the temperatures from going as low as they might 8 otherwise go, and the whole thing seemed to kind of wash 9 out with about the same risk.
10 For that reason we concluded that the B&W plants 11 should indeed have the same screening criteria applied to 12 them as other plants.
s, 13 MR. REED: Well, I know there has been a recent
( }
k" 14 SQUG event, and I heard some discussion, and maybe you 15 will have to tell me whether the pumps were stopped or 16 not.
17 MR. WOODS: Okay. You will recall a second ago 18 I said the significant events. A lot of the significant 19 events that have a lot of risk associated with them to 20 involve pump stoppage, and that makes my point. This did 21 not involve pump stoppage and it was not a significant 22 event. PTS was.
23 MR. REED: But there was cooldown.
24 MR. WOODS: There was cooldown to 386 degrees, 25 which is nowhere near significance for PTS purposes.
8
69
, 1 MR ., REED: All right. You are saying that the 2 cooldown has to be to 280 degrees.
3 MR. WOODS: It has to be lower than that, 4 considerably lower than that, even with the vessel at the 5 screening limit, which it was not. The vessel at the time 6 of that event was approximately 217 degrees RTNDT. The 7 screening limit is size 270. It would not have been 8 anywhere a problem even had the vessel been irradiated to 9 270, and it was even less of a problem at 217. We are 10 talking about many orders of magnitude.
11 In order to get at the risk, I have to look at 12 the more severe events, and a lot of them involve loss of
, 13 forced circulation. This one did not.
i i 14 MR. OKRENT: Could I ask whether anyone has 15 looked at the events that occurred at Rancho Seco, Oconee 16 and Crystal River, and all of them have had loss of 17 instrument, non-nuclear instrument power, to see how close is one was, in other words, what additional failure, if any, 19 there was, and I will use the word " plausible," might have 20 been dependent, or so forth, and could have gotten such a 21 reactor into trouble were it a highly irradiated vessel?
22 In other words, I am not talking about Rancho 23 Seco being in trouble from PTS from the 1977 event, if 24 that was the year. This was a new vessel. I am trying to 25 understand whether we had a precursor that was sort of one 8
70 l 1
(_; I step more from something that would be a real challenge or 2 whether in fact we we down a road where you weren't going 3 to get any further and you know it.
4 MR. WOODS: Well, obviously you weren't in that 5 latter case. You started down a road ---
6 MR. OKRENT: How about the Rancho Seco in 19777 7 MR. WOODS: 1978 is the last ---
8 MR. OKRENT: '78, whichever.
9 MR. WOODS: The light bulb incident.
10 MR. OKRENT: yes.
Il MR. WOODS: That one had already happened of 12 course when we recommended the screening limit that we now
I 14 reactor unless depressurization was automated to prevent 15 Pressurized thermal shock?
16 MR. WOODS: I have herd vaguely about it. I am 17 not what you would call ---
18 MR. SHEWMON: I don't think it would be possible 19 for them to have a pressurized thermal shock mostly 20 because they are buying a new vessel and if they have one 21 whose transition temperature will get up above 120F they 22 ought to shoot the people buying it. They are a competent 23 bunch, and I can't imagine they would do it.
24 MR. REED: Well, I can provide you, and I think 25 I can find it, with a piece of paper that says the British 8
J
76 m
t , I will have automated controls that will depressurize by 2 blowdown the primary system in order to prevent 3 pressurized thermal shock.
%e d MR. SHEWMON: But they also would not have a 5 vessel that would be subject to it even if they didn't.
6 MR. REED: Well, then why are they so worried to 7 put in automated depressurization to the primary?
8 MR. SHEWMON: I don't know. It is not clear 9 they want it only for PTS.
10 MR. WOODS: That sort of system has been 11 discussed as a possible corrective action since the very 12 earliest days of pressurized thermal shock, and we have
<, 13 not chosen to go that way, and the reason we haven't is 14 because you have to be awfully, awfully careful when you 15 put on such a system that you haven't created a worse 16 problem somewhere else than you have solved. We have 17 never been convinced that we knew how to do that. So we 18 haven't chosen to do that.
19 MR. REED: Keep working on it, because when you 20 find out it also works for decay heat removal, you will 21 want it.
22 MR. SHEWMON: Jesse.
23 MR. EBERSOLE: I just would like a 24 clarification. In your looking at therma) shock, do you 25 find the worse case is when the pumps are allowed to 8
77 1 continue to run or when they are stoppec and then you get 2 temperature profiles with large differences through the 3 temperature structure?
4 MR. WOODS: Well, like I said, we find an awful 5 lot of the risk is tied up with events where the pumps 6 have stopped and thereby allowing some kind of 7 stratification. But there are some events, secondary 8 system events where it is worse if you leave the pumps 9 running.
10 MR. EBERSOLE: Well would that be a main 11 steamline failure on the secondary side with full 12 feedwater flow?
13 MR. WOODS: Yes.
i i
\~' 14 MR. EBERSOLE: Is that about the worst one you 15 can get?
16 MR. WOODS: But, that is a leading question. It 17 varies so much from plant type to plant type and there is 18 just no general answer to that question.
19 MR. EBERSOLE: Well, I would think it would be 20 worse for B&W because you flood a superheated steam 21 section which is normally not wet and you get a vast 22 increase in cooling surface.
23 MR. WOODS: You get a change in area, yes, with 24 the OTSG that you don't get on the other vendors.
25 MR. EBERSOLE: So there is no clear picture as Ill
78 c_ _j i to whether the pump is on or off.
2 MR. WOODS: There is no clear picture. In fact, 3 I always carry with me a report that I think I have here 4 in the stack somewhere that Oak Ridge did as a result of 5 their three prototype studies that gets into all the 6 different factors that interact to cause one type of 7 transient to be worse on one plant under one condition, 8 because I know I can't answer that kind of question and I 9 would have to look it up for you.
10 MR. SHEWMON: Other questions on this one?
11 (No response.)
12 Okay. The second one had to do with how well
,- s 13 justified is the crack distribution used in developing the kl
"" 14 NRC position in PTS, and is there a sufficient base to e
15 justify the distribution used?
16 you can see the comment there.
17 Harold, would you care to comment on what 18 conclusions you came away with on that?
19 MR. ETHERINGTON: I might remind the committee 20 that at the January meeting I mentioned three points. I 21 was relaxed about the PTS rule, but I was concerned that 22 some waivers might be given without proper conservatism, 23 and I mentioned three respects in which I thought the PTS 24 was not overly conservative.
25 The first was that it used the Octavio crack
=
8
79
(_,! I distribution, which is based on rather flimoy grounds like 2 most crack distributions.
3 I did mention that in Europe there were other 4 distributions and mentioned the Marshall distribution 5 particularly.
6 The staff says that the Marshall distribution 7 included some vessels not of the quality of reactor 8 vessels. They thought that Octavio was kind of the best 9 they had. They agreed that it was the biggest uncertainty 10 in the whole business of PTS evaluation.
11 My other point about the crack was that there 12 was no allowance for crack growth. The Marshall
,, 13 distribution did allow for crack growth.
k '!
14 As the PTS stands, I believe there there is no 15 crack growth allowed for in the rule.
16 That was the first of three points.
17 Shall I pick up the others now?
18 MR. OKRENT: Is the answer that they gave, do 19 you think it disposes of your questions?
20 MR. ETHERINGTON: I was going to end up with a 21 recommendation that the committee ask to review the next 22 request for a waiver of the PTS rule.
1 23 MR. SHEWMON: This is the first time somebody 24 comes in having been caught by this trip wire.
25 MR. ETHERINGTON: Well, I think it is an 8
80 t s I unsatisfactory area, but I don't know how it can be 2 improved on.
3 MR. SHEWMON: None of us may be here when that 4 trip is there, but we will get it in the archives.
5 MR. MARK: Harold had two other points.
6 MR. SHEWMON: Does that cover the distribution 7 of the ---
8 MR. ETHERINGTON: That covers the crack 9 distribution.
10 MR. SHEWMON: Some of them come later. Okay.
Il The next one was throughout the transition 12 temperature range the staff appears to permit use of rs 13 unirradiated data from the upper shelf as its ceiling for i 1 iU 14 facture toughness. Is this so, and if so is this best 15 estimate, may it not be unconservative and, if so, by how 16 much?
17 Again, the staff has acknowledged that this is a 18 place where there is less data than they thought there 19 was, and had a program, or started a program to get 20 answers.
21 Would you summarize what you got from that, 22 Harold?
23 MR. ETHERINGTON: Yes. Again, I would remind 24 the committee of the issue. The irradiation does three 25 things. It raises the transition temperature, it lowers 8
81
_j 1 the upper shelf and it lowers the fracture toughness for 2 the Sharpe value throughout the transition range.
3 The staff had admitted to consider the 4 depression in the transition range, but I think the staff 5 is inclined to agree that this was an oversight. It is an 6 oversight, incidentally, that would transfer all of the 7 vessel low temperature studies, the heatup and cooldown 8 requirements, the pressurized thermal shock and the 9 inadvertent overpressurization.
10 I think that again there is an unconservatism 11 that we want to review during subsequent requests for 12 waivers from the rule.
13 MR. SHERMON: Okay.
( )
- ~' 14 MR. ETHERINGTON: There was one other point that 15 I made that is relatively unimportant. I mentioned that 16 the studies had not taken any account of the thermal 17 gradient in the vessel during operation. The inside of Ic the vessel is colder than the outside and therefore there 19 is a thermal stress on the inside of the vessel wall.
20 At the end of the day I thought I had 21 overestimated this, and I wrote a memorandum that was 22 transmitted to the staff the Monday of the next week 23 saying I had overestimated it. I guess that it might be 24 about a 50 degree difference through the wall, which again 25 I guess might be about a 10,000 psi stress.
8
82 7,
u_j 1 The staff made a calculation, or they dag it out 2 of the Westinghouse report, and I have overstated it by a 3 factor of about 2.
4 The actual thermal stress is about 5,000 psi on 5 the inside of the vessel, and if you had a three-inch deep 6 crack, that would give you a K-1 fracture intensity factor 7 of, oh, I suppose about 15. That is small, but it is not 8 really inconsequential and I thought probably ought to be 9 included in any future review.
10 MR. SHEWMON: Okay. I think that covers that 11 then. Thank you.
12 The third item had do with the effects of
~
13 reinserting water after partial core uncovery, and could
!""" l 14 this be a cause of pressurized thermal shock. This is one 15 that Ivan had brought up, and the position of the PTS 16 project or Roy Woods was that the partially uncovered core 17 in which you dump cold water is really a severe accident 18 question and was beyond the purview of their study.
19 I think we got a commitment out of some part of 20 the staff that they said the severe accident people would 21 look at this but the PTS people would not have it as part 22 of their purview.
23 MR. OKRENT: I have a problem with the answer in 24 the following way. Catton's point was that this seemed to 25 fallen into a crack in a sense. In other words, the PTS 8
83
/
, 1 people had given it to the severe accident people, but the 2 severe accident people weren't really looking at PTS.
3 MR. SHEWMON: The question here is what are the 4 consequences of you have got a dried core and you dump 5 water on it.
6 MR. OKRENT: Well, it could be partly dried 7 out. In other words, Three Mile Island was cooled down a and repressurized, and had that vessel been a highly 9 irradiated one with some flaws, it might have been 10 awkward, what Ivan is saying, you know, another it illustration.
12 MR. KERR: It certainly would have been 13 automatically depressurized, wouldn't it?
14 MR. OKRENT: That is right, and Glenn might have 15 been real happy about that. But it is not clear to me 16 from what I read in the severe accident area that they are 17 really thinking about that aspect of the problem.
18 MR. SHEWMON: Well, a different question is do 19 you feel this is a high enough probability event so that 20 it would change the PTS result?
21 MR. OKRENT: I don't know. People are saying, 22 again this was Ivan's point, so I don't claim originality, 23 but I concur, people are saying that the chance of damage 24 and recovery from damage is larger from the chance of 25 large-scale core melt, you know. There is this factor of 8
84
.. j 1 10 people were talking about.
2 If that is the case in fact, and if you accept 3 that, and the chance of large-scale core melt, and let's 4 take a number, 10 to the minus 4, and you believe this 5 factor of 10, that I think it is an exaggeration, you 6 would be closer to 10 to the minus 3 for getting to the 7 possibility.
8 Now you need lots of things to be present to be 9 in trouble. You need a high irradiated vessel and so 10 forth and so on.
11 But it is not obvious to me that you are at a 10 12 to the minus 7 situation or a five times 10 to the 6
,, 13 situation, which the staff has in their SRP with which I
+
'"S 14 disagree. In other words, they say with the situation of 15 five times 10 to the minus 6 or smaller you can ignore it, 16 and I think that is just a little bit big to be talking 17 about something that involves pressure vessel failure.
18 Well, you know, we can all bandy about numbers 19 and you can probably bandy them about with more authority 20 than I can. If we say that an overheated core is 10 to 21 the minus 4, then the 10 to the minus 1 on recover would 22 get you to 10 to the minus 5 as a severe one, but it still 23 comes up to the edge of what he is talking about here.
24 MR. OKRENT: If the overheated core is 10 to the 25 minus 4, and you are saying core melt is 10 to the minus l
l 1
l
85 1 5, you know, earlier this morning when people were talking 2 about safety goals, they were saying, oh, I want to 3 distinguish between large-scale core melt and this 4 recovery accident, and one is that large-scale core melt 5 is much less likely.
6 MR. SHEWMON: I agree, but I don't happen to 7 think that the large-scale core melt on the floor is 10 to a the minus 4. From all the calculations I can see, if they 9 get the core on recovery, it is a core melt because they 10 don't know what to do with it from there.
11 MR. OKRENT: Well, in any event, the logic is 12 that if people are right that recovery is likely let's say 7~ 13 half of time, and if you calculate -- well, I will just be 14 repeating myself.
15 I don't think the staff has disposed of the 16 answer on a probabilistic basis. I would be happy td'have 17 them do it, but at least not here they haven't, and I 18 don't know how the severe accident people are treating 19 it. So I think Ivan's question about it being in a crack 20 remains.
21 MR. SHEWMON: He is putting words in your 22 mouth. Do you want to speak for yourself?
23 MR. WOODS: Yes. I believe you are correct whed 24 you weren't sure that this was being handled by the severe 25 accident people. At the time when this whole thing 8
86
_; I started because of Professor Theofanous' paper, it 2 wasn't. We hadn't thought of it. But we made them aware 3 of it and they are in fact now considering it.
4 Now the first part of their consideration will 5 try to be to decide how frequent this is and whether or 6 not it warrants consideration further than that. But it 7 is definitely on the program of the severe accident people 8 and it will be considered. I can't tell you at this point 9 how thoroughly it will be considered because it depends on 10 the output of the first steps.
11 MR. OKRENT: All right. It seems to me what 12 would be helpful is for somebody to make a best estimate,
, - , 13 however crude, of what the frequency range for this event
( !
I' 2 14 might be.
15 MR. WOODS: That is exactly the first step that 16 they intend to do, the severe accident people.
17 MR. OKRENT: If they do come up with numbers 18 that are larger than your screening criteria, it seems to 19 me it moves back into PTS consideration and you have to 20 rethink where you are.
21 MR. WOODS: PTS then becomes one of the 22 considerations to be worried with along with the 23 consideration of trying not to melt the core by 24 overheating it as you are trying to recover from a severe 25 accident.
8
87 s / 1 MR. OKRENT: Exactly. Okay. Well let me just 2 leave it at that.
3 MR. SHENMON: Fine.
4 MR. OKRENT: Are you talking about six months or 5 six years or what scale for this estimate?
6 MR. WOODS: The severe accident people were here 7 before and they aren't here today. I can't answer for '.-
8 somebody else.
9 MR. OKRENT: Okay.
10 MR. SHEWMON: The next question. Are there any 11 steam generator overfill scenarios which the staff 12 considers significant for PTS 7 13 MR. WOODS: I am sorry. Where are we? Oh, 14 okay.
15 MR. SHEWMON: Yes, there are steam generator 16 overfill scenarios that are important to PTS, it says 17 here.
18 MR. WOODS: That is right, and we had Carl 19 Johnson here to answer that question very thoroughly from 20 the prototype analyses that were done at Oak Ridge, and I 21 believe he discussed which ones were looked at and how 22 much they contributed to the risk, and unfortunately I am 23 not prepared to repeat that today.
24 MR. SHEWMON: Okay. There is something there.
25 MR. OKRENT: I can read it then. If somebody 8
88
,-q m_; I will get me the pages of the transcript, I will read it.
2 MR. SHEWMON: Are there any reactors for which 3 the data and chemical composition of critical welds is not 4 determined, not well determined and, if so, how is a 5 judgment made? Is the difference between the composition 6 accepted and the worst possible significant and, if so, 7 how much less likely must the worst possible be and how is 8 this judgment made?
9 And there is a page of comments there talking to about the different possibilities and how they have 11 handled them.
12 MR. OKRENT: I read the answer, and I guess what 7- , 13 I had in mind was my understanding of the available 14 information on a reactor that was looked at in this sense 15 recently where there were not samples from the actual 16 weld. There were samples ---
17 MR. SHEWMON: There were records showing the la same heat of wire and flux had been used for welds they 19 could check.
20 MR. OKRENT: Yes, and they had other samples, 21 which if I understand correctly that some were -- there 22 were maybe three and two were good and one was not so 23 good, or something like this. That is my recollection.
24 What I am trying to get at is what degree of 25 confidence, if I can use that bad word, does one want to 8
89 j 1 have with regard to this part of the problem, and is one 2 getting it from the approach and the informatjon that was 3 available to that reactor.
4 What I have now is, as I understand it, the 5 judgment is that the ---
6 MR. SHEWMON: Randall's judgment was that it was 7 adequate.
8 MR. OKRENT: Yes, but not, if I remember 9 correctly, not all of the weld samples that they had were 10 low in copper or whichever was the bad contaminant. It 11 wasn't clear to me that in a place like this where you are 12 looking for rather low probabilities of failure, I didn't s ;3 know what assurance one needed in this area. That is only l
(\'
14 part of the overall problem. But what assurance you 15 needed in this area and how they had garnered that 16 assurance, or could they do anything ---
17 MR. SHEWMON: What they had done to begin was to la assume this was the worst conceivable composition and l 19 transition teinperature, and that is what got Robinson in 20 trouble in the first place. And given that albatross i
21 around their neck, they then went out with the idea of l 22 trying to get some evidence that indeed this wasn't the 23 highest copper weld ever recorded any place by anybody, 24 but was something better because of the material they new 25 they were using at that time.
E
90 g-. -q
(_ / 1 So what they did was having just to go back and 2 look at record on what heats were being used then and what 3 fluxes were. So the weld wasn't taken out on that day 4 from that time. The heats of these wires are pretty big.
5 So there is a lot of continuity with them.
6 Harry looked at this question of why there was 7 this one number farther out and concluded that the staff's 8 explanation was reasonable.
9 I think any way you look at that data it brings 10 you up well above this worst one anybody could find any 11 basis for, which is what the staff had had them assuming 12 to begin with, and that was where H.B. Robinson was in s, 13 trouble. Once you get back to something with lower copper i 1 2J 14 than that, then their trip point moves off some place 15 beyond or certainly to the end.
16 MR. OKRENT: Does one need to be 90 percent 17 confident that you are not off by some factor in the wrong 18 direction or only 50 percent confident or 99.9 percent 19 confident concerning the amount of copper and whatever it 20 is, nickel?
21 MR. SHEWMON: There simply is no basis for that 22 kind of number. We could send somebody out to do a PRA 23 and that might comfort you, but that is simply not the way 24 things are done in that area. You end up with the 25 uniformity of the weld wire and the uniformity of the 8
91 l
) I welds that come from it, and it is possible that somebody 2 could go back and get all the chemical analyses that ever 3 came out of the heat and put them end to end and do sort 4 of a confidence level of the kind you are talking about, 5 but nobody has, and in general the people who work in that 6 field don't' feel it is necessary or fruitful.
7 Well, again, they do use a quantitative a screening criteria. If they only need to have a two out 9 of three or even a nine out of ten here, that would be one 10 thing. If they really think that they need to have this 11 number, like a 99.9, then they probably haven't 12 demonstrated ---
13 MR. SHEWMON: They don't.
( :
14 MR. OKRENT: Well, I don't have a feel for what 15 confidence they need in this area. It may be that there 16 are some statistics if you just look at the variations 17 that occurred on other vessels where they took samples ---
18 MR. SHEWMON: There are statistics on variations 19 through a heated wire, and that is very small.
20 MR. OKRENT: All right, that is through a heated 21 wire, but there may be situations where it was thought 22 that there was a heated wire, and yet there were 23 differenceo on other vessels. I don't know.
24 MR. SHEWMON: That is what they did this test 25 top and side for, to see if indeed if that did have the 8
92 7-7 j 1 composition that looked to be safe, and if it did, then 2 they would assume that the heat -- they had records to 3 show that the wires used -- it was the same wire or heat 4 of wire used in the two welds and therefore they knew it 5 wasn't what the staff had assumed to begin with.
6 MR. OKRENT: Well, I don't think there is a zero 7 chance that the amount of copper is higher than they are 8 now using. It may be very small or it may be 9 significant. I just can't tell, that is all, from what I 10 read. I am unable to tell whether the uncertainty is 11 significant. So I don't know how well they need to ---
12 MR. SHEWMON: I don't know what the uncertainty
,_ 13 is you are talking about, whether you are talking about l
1 A e" 14 the uncertainty from one end of a wire to the other end, 15 or whether the uncertainty is when they have record s that 16 say the same wire was used middle and top that it was 17 indeed the same wire. Well, either of those could be 18 uncertain, but the chances are extremely low, like zero, 19 or something less or bigger.
20 MR. OKRENT: Well, I am interested really in l 21 what is in the weld material, and if the uncertainty in 22 that, you know, is really zero, and the convincing case 23 has been made. l 24 MR. SHEWMON: The uncertainty along the wire is 25 not zero.
Ill I
t 93 nj 1 MR. OKRENT: All right. Well, then you say it 2 is bounded narrowly. If that is the case, and one can 3 conclude that there is just no chance that something else 4 is in the weld material and there is no chance of having 5 higher, significantly higher, I am willing to be 6 convinced, and don't misunderstand me. What I read seemed 7 a little bit ---
8 MR. SHEWMON: Randall has gone through a lot of 9 this, especially with the revision of Reg. Guide 1.99 on 10 what we allow people for estimating shifts in anneal 11 ductility temperature. They have standard deviations on 12 what kinds of variations come in a weld, or along a weld
- 13 and out of a given heat of wire and flux, and those things
/ ;
14 were factored into the estimates of the anneal ductility 15 temperature that were in Reg. Guide 1.99 and also in the 16 PTS.
17 And if they felt there was an argument for 18 shifting the mean, I am sure they still took those two 19 sigma or three sigma maybe, and I suspect they used two, 20 on top of it to get what looked like a conservative copper 21 content.
22 MR. WOODS: I believe Dr. Okrent needs to talk 23 to Dr. Randall.
24 MR. OKRENT: I would prefer, if there is 25 something in writing that already exists that I can read 8
94 rq ts. ) I that will convince me or'give me information, that would 2 be easier.
3 MR. WOODS: I know that from personal experience 3
4 I had the same sort of questions you have, how could we be 5 using .35 copper and now all of the sudden we are using 6 .22 copper, and I have discussed it rather extensively 7 wjth Niel Randall. Unfortunately, Dr. Randall takes 8 alternate Friday's off on the schedule, and this was two 9
weeks ago and it would have cost me my life to bring him 10 down here today.
11 (Laughter.) '
12 MR. OKRENT: I assume that -- well, maybe I am g, 13 wrong, but that is a non-trivial difference of .35 to .22.
\2 14 MR. WOODS: It is a very non-trivial difference.
15 MR. OKRENT: So, therefore, I would assume that 16 there is more somewhere in writing than the usual three 17 sentences that one sees in an SSER.
18 MR. WOODS: Therefore, staff believes they have 19 good reason to ae64te the new value instead of the old 20 value, and 7 OP..nto document that for you, but Dr. Randall 21 can.
22 MR. OKRENT: If there is such documentation that 23 he has, and I can get a copy, I would be pleased ---
24 MR. SHEWMON: Now we asked Perry to look at 25 this, and Perry assembled a documentation and gave us a S
95
.( ~
R3 1 report. Have you read that lately?
2 MR. OKRENT: I must confess it is some time 3 since that came out. So I have not read it lately. It is 4 my fault, and maybe I can get this from Perry.
5 MR. SHEWMON: Yes, I think that would be a good 6 place to start.
I 7 MR. OKRENT: All right, fine. Maybe Al can s arrange to get me the relevant material.
. 9 MR. IGNE: (Nodding affirmatively.)
10 HMR . OKRENT: Fine.
11 MR. ETHERINGTON: I am clear what copper is
, 12 reported. Is it reported in the wire after the copper
- 13 plating, is it the mill composition or ---
14 MR. SHEWMON: It is the weld composition.
15 MR. ETHERINGTON: Oh, it is the weld 16 composition. Okay.
17 MR. SHEWMON: And the argument is that it is 18 uniform once they use the same wire and the same heat 19 within the two sigma that he used.
20 What is the expected consequence of a 21 throughwall crack, and what is the likelihood of core melt 22 late containment failure and early containment failure?
96 u; 1 just not part of their study.
2 MR. OKRENT: Well, let me say why I posed the 3 question.
4 When I looked at the -- and I don't know whether 5 it is a draft or if it is an actual SRP section, I can't 6 remember any more -- it says that when you are considering 7 PTS, if the scenario has an estimated frequency of less 8 than 5 times 10 to the minus 6 of causing a throughwall 9 crack that you don't have to worry about it. In other 10 words, they are cut off at 5 times 10 to the minus 6. Am 11 I wrong?
12 MR. WOODS: We don't have to worry about it?
, . , 13 MR. OKRENT: My recollection, and correct me if I
' ") 14 I am wrong, is that 5 times 10 to the minus 6 is the 15 frequency.
cw 16 MR. WOODS: That frequency is meant to apply to 17 the sum total composite of all the transients that might 18 occur.
19 MR. OKRENT: All right, let me leave it for all 20 the transients. No, not transient with occur, but 21 transients which occur and lead to throughwall crack. The 22 number of transients that occur is much more frequent.
23 MR. WOODS: The precursors don't count, but only 24 precursors severe enough so you predict a throughwall 25 crack.
8
97 r~s
'k ) 1 MR. OKRENT: So, therefore, in effect the cut 2 off for a throughwall crack is 5 times 10 to the minus 6, 3 and if you got 4 times 10 to the minus 6 for the total, 4 you could go on irradiating for a while. Let me put it 1
)
5 that way, as I understand the SRP.
6 MR. WOODS: Okay>
7 MR. OKRENT: All right. Now if this has a high a likelihood, let me say two chances in three of leading to 9 early containment failure, it is a non-trivial event.
10 MR. WOODS: Yes, we know that, and in fact we 11 did as part of PTS look at the consequences of the 12 throughwall crack.
13 MR. OKRENT: This is why I asked the question.
14 MR. WOODS: I know, and the answer to it is is there is a paper that was presented by Rich Barrett and Ed 16 Strom at the last water reactor safety meeting in 17 Gaithersburg which I think we passed out at the 18 subcommittee meeting which you need a copy of.
19 MR. OKRENT: Okay. I will be glad to read it.
20 Has the paper been peer reviewed and everything?
21 MR. WOODS: It was presented, and I guess that 22 gets it some kind of review.
23 MR. OKRENT: Not automatically.
24 MR. WOODS: Well, I an not sure exactly what you 25 mean, but it certainly has been made public and 8
98 U
L; I discussed. Al Igne had a copy.
2 MR. SHEWMON: Well, actually there is a better 3 answer I guess in what is'in front of you there on page 16 4 than what I gave. It gives the title and some of the 5 results.
6 MR. WOODS: And that paper, incidentally, was 7 based largely on a contractor that NRR had at Pacific 4
8 Northwest Labs.
4 9 MR. OKRENT: Let me say what I mean by peer 10 review. The NRC says that in fact they are going to have 11 PRA's, peer review, but I think they mean PRAs that are 12 done by licensees. I think equally well that when you
, 13 have something done for the NRC by a contractor, for O 14 example, or even within the NRC there needs to be a 15 mechanism for peer review. That is what I meant. And 16 presenting it at a meeting is useful, but it is, you know, 17 and you know as well as I do that it takes substantial 18 _ effort to really peer review something like that. You 19 don't do it by listening to a 30-minute presentation.
20 That is the sense of the question, but I don't 21 want to pursue it further and I will read the paper if I 22 have it or someone gets it'to me.
23 MR. REED: What is the status of this letter tc 24 Stello? Is it going to be sent, polished and sent or 25 what?
111
s 99
( ) 1 MR. SHEWMON: I don't know what letter you are 2 talking about.
i 3 MR. REED: This proposed letter draft one that 4 Okrent wrote.
5 MR. OKRENT: That was ---
t 6 MR. REED: A long time ago?
7 MR. OKRENT: No. Last month I brought in this 8 draft, which included Harold's questions, Ivan's questions 9 and my questions, and we didn't take action on it, but 10 Paul used it as a point of discussion. So it has served '
11 its purpose already. It just intended to serve as a point 12 of trying to get information.
13 MR. WOODS: We used that as his list of 14 questions to would be discussed at the subcommittee 15 meeting.
16 MR. OKRENT: Yes. So there is no point in 17 sending it at this stage.
i 18 MR. SHEWMON: That is what we have been going 19 over.
20 MR. REED: Then can I leave a question for you.
21 I would have added another one, No. 8.
22 (Laughter.)
23 MR. WOODS: Talk to my management.
24 MR. OKRENT: Sorry if I missed you, Glenn.
, 25 MR. REED: In view of the complexities of
100 y
u, a 1 accident scenarios and the response of various equipment 2 in brackets, recirculation pumps, high pressure safety 3 injection pumps, charging pumps, auxiliary and main boiler 4 feed pumps, PORVs, et cetera, close brackets, in view of 5 the response of that various equipment to provide fairly 6 rapid temperature reduction and high pressurization, we 7 request that the NRC staff review the reasons behind the 8 recent Central Electricity Generation Board decision to 9 provide an automatic and integrated control system to 10 depressurize by primary blowdown in order to deal with 11 PTS.
12 MR. SHEWMON: The next time we write a letter on
,7 13 PTS we will have it there. Is that something you wanted 14 the staff to common on ---
15 MR. REED: Well, I would like to have them think 16 about it.
17 MR. WOODS: We have thought about it for years, 18 and we concluded for what I believe are good reasons that 19 we did not wish to pursue it that way. I don't know if it 20 is documented or not.
21 MR. REED: You haven't thought about it since 22 the CEGV thing because you didn't even know about it.
23 MR. SHEWMON: He has certainly known about that 24 as a way to copy with PTS. He said that earlier, and I 25 suspect he means it.
8
101
/
(]_/ 1 MR. REED: Well, I mean the reasons behind their 2 solution.
3 MR. MINNERS: Those are very difficult. We have 4 tried to do that in many cases and, you know, you don't 5 get very good answers. You really don't get very good 4
6 answers from that. They have the same difficulty in 7 making decisions about things and they have political e pressures on them like we do. So I. don't see the sense.
9 They made their decision, and I am sure they stated the 10 reasons.
i 11 MR. REED: You have to jump for this. You could 12 go to Great Britain on a trip.
13 (Laughter.)
- 14 MR. OKRENT: I will try to write someone in j 15 England, since the staff doesn't seem to want to explore i
16 it, and see whether it is so difficult to get the logic
! 17 for the position.
4 18 MR. WOODS: I of course agree with my management l 19 here. I work for this gentleman. But I. nave tried to get 1 20 information concerning operating events and so forth, j 21 something that simple from the foreign governments, and it 22 is not just worth the trouble. You don't get something 23 that is useful to the point of the effort that you 24 expended to get it.
25 MR. SHEWMON: Mr. Chairman,-I think we have
- lli
102 y- .
L_) I worked this ovel- and gone through the questions, and I 2 would turn it back to you.
3 MR. WARD: Okay. Thank you. Let's go right to d our next topic. j 5 MR. OKRENT: I am sorry to hear the staff give 6 so negative an attitude about trying to 2 ear.n why other 7 people in other countries make safety decisions or what 8 they have learned fi-om operating experience.
9 MR. KERR: Now, Dave, he didn't. say that. He lo said he had tried to get information and he had had great 11 difficulty getting it.
12 MR. OKRENT: But he also said it wasn't worth
<s 13 the effort, and my experience is ---
14 M.R . KERR: But you are not an employee of the 15 U.S. Government dealing in a formal way with the 16 government over there, and you may have channels that he 17 doesn't have.
18 MR. SHEWMON: What is the next topic?
19 MR. WARD: Maybe we can look to the U.K. for 20 operating experience with LWR's.
21 MR. KERR: I need to make a phone call, and it 22 will take me not over five minutes. Is it possible to 23 have a five-minute break before we go ---
24 MR. WARD: All right, let's take a five-minute 25 break.
lli t
I
103 1 (Brief recess.)
2 MR. WARD: The next topic is a discussion of the 3 severe accident policy.
4 Dr. Kerr.
5 MR. KERR: Under Tab 12, if you have not already 6 looked, there is information bearing on the topic to be 7 discussed, including some draft set of minutes of a 8 meeting on February 24th of the Class 9 Subcommittee and 9 representatives from NRR who discussed an implementation to plan for the severe accident plan.
11 The plan is divided into three general areas 12 that have to do with systematic examination of individual 13 plants, that have to do with the procedures to be used on 14 new plants that will involve PRA and some non-15 probabilistic criteria, and possible changes in rules or je regulations that may be based on new information developed 17 as a result of the source term research.
18 Much of what we talked about in the meeting, 19 because I think it farther developed, had to do with the s
20 first of these, the systematic evaluation for examination 21 of individual plants that is being developed by NRR in 22 cooperation the the IDCOR program people.
23 The staff would like us to com4nent on what they 24 plan to do at this point as far we can with the existing 25 information so that they can get some idea of whether we t -
1 104 F7 l_) I approve, if we disapprove or have additional suggestions, 2 they would like to hear them.
3 I ask for a somewhat abbreviated presentation d
which may constrict the people making the presentation too 5 much, but for one which would perhaps take not more than 6 an hour if they were not interrupted by questions from 7 this group.
8 There were other people, or at least one other 9 He is not present here today. So I won't person present.
10 ask for his comments. I will turn things over to Mr.
II Rosztoczy at this point and ask him to orchestrate the 12 further presentations.
13 MR. ROSZTOCZY: Thank you, Mr. Chairman.
14 My name is Zoltan Rosztoczy from the NRR staff.
15 We also have with us today Frank Coffman from the NRR
,g 16 staff and Len Soffer who are responsible for various 17 portions of this program, and we have with us Jocelyn 18 Mitchell from the Research Office for any support what we 19 might need from that direction.
20 (Slide.)
21 My subject today is a brief summary of our 22 implementation program on the severe accident policy and 23 the regulatory use of the source term.
24 Before we get to the implementation plan, let me 25 first just bring to your attention the basic action items l
Ill l
-. _ _ ~ _ _ - - _ , - - - - - _ _ _ - . _ _ _ - - _ . _ _ . _ , - . -- - - . _ - . _ . _ - - , _ _ _ _ .
105 fs Is.) I in the Commission's policy statement.
2 (Slide.)
3 The Commission issued its policy statement on 4 severe accidents last August, and policy statement is 5 basically organized in two parts, one for new applications 6 and the other part for existing plants.
7 Under the new application part there are 8 basically two action items for the staff. One of them is 1 9 to issue guidance on the role of PRAs. A rule called the 10 CP rule is already out'and already in effect which Il requires that new applicants prepare a PRA as part of 12 their application for licensing a nuclear power plant.
i 13 However, there is very little guidance of what 14 is the extent of the PRA what they have to perform and t
i 15 what criteria is it going to be compared against.
1 16 So the staff's assignment is to prepare this 17 guidance and also prepare performance criteria for la containment systems if such a criteria is needed. So as 19 part of the program we will look at the need for 20 containment criteria and then also prepare the basis of 21 such a need for criteria.
22 The second part of tne policy statement 23 discusses existing plants. Relative to the existing 24 plants, the Commission emphasized that they do not see any 25 reason at the present time for immediate action. However, Ill 1
_..y.,.-r, , e . _ - , - ,p ,. . .-w-.. . - - - , , ,,p, .-.,_.~~,..,....-+--,e-.-ym.,-e., .c---c.r_,
106 r9 kj 1 systematic examination of the existing plants for severe 2 accident vulnerabilities is needed.
3 So part of our program is to prepare the t
d guidance that needs to be given to utilities who then will 5 perform the systematic examination.
6 The second part is that if we find any changes, 7 any plant dependent weaknesses or vulnerabilities, then we 8 should use the backfit rule to decide whether a backfit is ,
! 9
.needed and whether a backfit should be required.
10 Also mentioned in the policy statement is the 11 question of any generic type of changes, and the policy 12 statement indicates that if there is a need for that type
- 13 of a change, then that should be accomplished through the
"^ 14 normal process through rulemaking.
15 So with this much of a background on the policy 16 statement, the staff has prepared a plan for the 17 implementation of the policy statement.
18 The implementation plan is arranged in three 19 elements, as was mentioned by Dr. Kerr.
20 (Slide.)
21 The first element addresses the existing plants, l 22 the second element is basically for the new plant 23 applications and the third element addresses changes in 24 rules and regulatory practice.
< 25 In terms of the first element, we intend to J
i-107 m
1 develop guidelines and also criteria that the examination l 2 can be measured against for the examination of individual i
3 plants.
I l-4 The methodology that will be used for the l
l 5 examination is being prepared by the industry, being 6 developed by the industry under the IDCOR program.
7 We are supposed to receive the IDCOR submittal 8 of their proposed methodology within a month or two, j
9 together with a few sample applications. Then we are 10 going to review it. By that time we hope to have some 11 strawman guides at least for the first plant, and we will 12 compare it against the strawman guidelines and see if they 13 fit together and see if the whole package together l-l O 14 provides all the guidance that the utilities will need to 15 perform this examination.
l 16 MR. WARD: Zoltan, can I ask a question at this l
l 17 point. You refer a number of times to the examination of 18 plants. To what extent is this -- and maybe you are going l 19 to tell us more about this -- but to what extent is this l
20 really an evaluation of plants, actual plants, and to what l
21 extent is it really an examination of the plant designs?
l 22 Say if the methodology is primarily some variant j
- l l 23 on PRA, I guess I would maintain that it is really an i
i 24 examination of plant designs and not of the plant if I )
25 define the plant as consisting of the actual equipment llI I l
1
108 I
t_s that exists at the plant and the actual organization and 2 the body of personnel who operate the plant.
3 This examination is expected to MR. ROSZTOCZY:
4 So it will include some include both of those.
5 examination of the plant design and some examination of l 6 the plant exactly as it is.
7 The IDCOR methodology includes a plant walkdown, 8 and part of it is to see the important thing what we are 9
relying on, what has been identified based on the review 10 of the design, that those things are really there and they 11 are there in such a fashion that they can perform the 12 function when it is expected that it will be performed and yy I3 so forth for those severe accident sequences where they do
( '" l 14 depend on those functions.
15 It is also related and directed rather strongly 16 to operator actions. So it includes, for example, time I7 and space studies whether an operator can perform a IB certain function what he is asked to perform. For 19 example, if containment venting is an issue for a given 20 sequence and whether the means, including the time 21 available, is there for the operator to perform this l
22 function, and then specifically oriented to whether the 23 equipment would be operable under the conditions.
24 MR. WARD: But does it look at the performance 25 of the equipment that is actually in that plant, for 8
1
l l
i
, 109 i
Ts 1 example, and does it look at test performance, failure 2 rates and so forth for the equipment in that plant? As 3 far as personnel performance, does it look at the 4 performance of the actual crews that man the the plant, or 5 is it some kind of a more hypothetical evaluation?
6 MR. ROSZTOCZY: In the case of an actual plant, 7 it would be looking at what actually is there and whether 8 it would perform. Now how do you decide that that 9 equipment would perform, in most cases we are talking 10 about rather adverse conditions, conditions that the 31 equipment was never exposed to, a possible exception of 12 the qualifying tests that it went through for the normal 13 design basis events.
14 So one item that is being looked at is to see 15 whether the environmental conditions during any of these 16 sequences during the time period that the equipment is 17 expected to perform, whether those conditions are more la severe than the qualification conditions and if they are 19 more severe than what needs to be done to assure that the 20 equipment will perform.
21 MR. MICHELSON: I am wondering if you could tell 22 me what accidents you now look at, keeping in mind that we 23 discussed this morning ---
24 MR. KERR: I would urge that you let him get 25 farther along with his presentation.
8
110 t J I MR. MICHELSON: Oh, okay. He is going to tell 2 us later. All right. Thank you.
3 Can I pursue a little bit what Mr.
MR. OKRENT:
4 Ward was asking about. In the first place, is there a 5 draft methodology ---
6 MR. KERR: Again, I really believe that if you 7 will give him a little time your question---
8 MR. OKRENT: Well, let me ask the question. Is 9
there a draft methodology that I can get to look at that 10 relates to the first two bullets under 17 II MR. ROSZTOCZY: The first bullet under the 12 existing plant IDCOR methodology, there is a draft
- 13 methodology, but we don't have it yet. This draft 14 methodology was given to the owners of seven plants. They 15 are using it ---
16 MR. OKRENT: The' answer is no then.
I7 MR. ROSZTOCZY: Since we got this far, let me 18 just give two sentences more.
19 It was given to seven plants, and those seven !
20 l plants are now using it. Information is being fed back to -
21 IDCOR, and based on that it will be finalized and the 22 final methodology is submitted to us within a month or so.
23 MR. OKRENT: And the same goes for guidelines 24 and criteria?
25 MR. ROSZTOCZY: Guidelines and criteria are 8
111 I being developed by the staff, by us. We are in the 2 process of developing those, and for the very first plant, 3 which happens to be a Peach Bottom plant, some preliminary 4 guidelines for the very first plant will be available 5 within two months.
6 MR. KERR: Don't hesitate. Keep going.
7 (Laughter.)
8 MR. ROSZTOCZY: So this was the existing plant 9 part. Then for the new plants, basically what we are 10 looking for is that we have various requirements presently 11 available which really go beyond the classical design 12 basis, for example, the ATHS rule or resolution of the
,e 13 station blackout problem.
14 We are going to look at all the available is deterministic requirements and see how that should be fed 16 into the evaluation of PRA results and what additional 17 requirements are needed and what is the minimum content 18 that the PRA should address and what should be the extent, 17 the minimum extent that the PRA should cover for severe 20 accident sequences and what depths should it go into.
21 The third area is the question of changes in 22 rules and regulatory practice. Regulatory practice here 23 means all the other regulatory items what we put out like 24 regulatory guides, standard review plans and so on.
25 We cut this into two separate areas. You have 8
112 I probably noted that the title of my discussion today was 2 the implementation of the severe accident policy statement 3 and source term related regulatory changes. We are doing 4
the two together because they are so closely intertwined 5 that we think it is better if it is done as one single 6 program.
7 Under the changes we will have a look at 8 separately the source term related changes and then any 9 other changes that might come up from the severe accident 10 program. And I will come back later to describe what are Il the source term related changes and what is the schedule l 12 that we are working on to try to accomplish this. '
13 (Slide.)
14 One question what we have to always answer is 15 what are we trying to accomplish by doing this work. !
16 Atter everything is done and we have done our work and the 17 utilities have done their examination of the individual f 18 plants, then what have we accomplished.
19 The basic goal is to identify plant specific 20 vulnerabilities, what plants might have severe accidents.
21 The time when these plants were designed there were no 22 requirements for severe accidents. So they didn't have to 23 design the plants for severe accidents.
24 Nevertheless, the available margin in the plants 25 in many cases can cover a severe accident, but there,vould 8 -
113 O
'Sss) I be some vulnerabilities and some weak points that need
, 2 additional attention.
3 So these weaknesses, these vulnerabilities will 4 be identified as part of the program, and if such 5 vulnerabilities are identified, then an appropriate fix
+
6 using the backfit rule is going to be selected correction.
7 The second item is that if we find some generic 8 vulnerabilities, something at not one plant but a group of 9 plants, then obviously we are going to address those, and 10 then we are going to use our normal procedures for that 11 which is rulemaking.
4 12 The third item what we hope to accomplish is 13 that by doing all of this we hope that to understand how 14 severe accidents can affect a plant and what features a 15 plant ought to have in order to be safe for a severe i'
16 accidents or relatively frequent severe accidents.
17 We will provide additional information for the l 18 designers of future plants, and hopefully they can factor 19 this into future designs and as such the future d,esigns 20 can be more forgiving in terms of severe accidents.
21 MR. WARD: Zoltan, will you learn anything about 22 the operator training as a possible serious cause of a i 23 plant being an outlier?
24 MR. ROSZTOCZY: Yes. One of the major goals is 25 to identify the operator training that we feel is
!Ill f
114 k_) I necessary in order to have the operators ready to handle 2 severe accidents should a severe accident develop 3 somewhere.
4 MR. WARD: Will the methodology permit you to 5
identify a plant where the operator training is poor?
6 MR. ROSZTOCZY: No. The purpose of the 7 methodology is to identify in which areas is there a need 8 for operator training in the severe accident arena as 9
opposed to the design basis. The operators have been 10 trained for design basis accidents. They have not been Il trained for severe accidents.
12 So the question is how should the operator
< w, 13 training being expanded so as to cover severe accidents
(
)
14 and what are the important things that should be included 15 in operator training. But we are not going to set up an 16 operator training program. That would be left to some 17 future programs. Therefore, we certainly wouldn't measure la the effectiveness of operator ---
19 MR. WARD: No, no. I mean the purpose of the 20 program is to identify outliers, weak plants.
21 MR. ROSZTOCZY: Yes.
22 MR. WARD: And from what I hear, the program is 23 really directed toward identifying weak designs of plants l l
24 pretty much period. l 25 MR. REED: I would be serious at least with !
8 l
l
115 m/ I respect to the operator training that you are thinking of l
2 to just take the NASA example. Obviously the astronauts, 3 there is you can do about it, but the design was flawed.
4 So you have to work with the design and the procedures.
5 MR WARD: That may be true, and if we i
6 absolutely know that, this sort of procedure is probably 7 fine. I don't absolutely know that.
8 MR. REED: I was being facetious and trying to 9 make a point, to look for the design vulnerabilities.
10 MR. ROSZTOCZY: We will address questions of II what operator action is needed in order to either prevent 4
12 seve're accidents or to mitigate the consequences of severe
, 13 accidents. And if we find that there are some operator 14 actions which would be very useful for either of these 15 purposes, we will identify those. We will also check 16 whether those are included in the current training of the l 17 operators, and if it is not included irt the current 18 training, then we will make a recommendation to include 19 it.
20 MR. OKRENT: I have two different questions.
21 First, if I understood your answer to my earl.ier j 22 questions, they are going ahead and analyzing seven plants I
- 23 and the staff doesn't even have a draft of the methodology 24 that they are using. That is what you told me.
25 MR. ROSZTOCZY: The staff has received two Ill
116 t O kJ I briefings on the methodology, but does not have a draft of 2 the methodology.
3 Let me maybe add so gou understand a little d
better. The methodology is set up as a question / answer 5 type of methodology. So it is poses a question and then 6 asks the utility to respond to this question, and :
7 depending on what the response is, it tells us where to go
- 8 next. So it is a big book for PWRs and a big book for 9
BWRs. We do not have that book yet, that is correct.
10 j MR. OKRENT: I must say I am unbelieving, and I 11
~will leave it at that.
12 The other thing is I would like to pursue Mr.
13 Ward's point specifically in the following way. After the 14 recent Davis-Besse incident I believe there were 15 considerable changes made in management and things like 16 this at Davis-Besse.
17 i Apparently there were some deficiencies that 18 existed and improvements were needed. Is there some way 19 in which the effort that you are discussing will uncover 20 such deficiencies if they are major? Is that a question 21 of interest?
22 MR. ROSZTOCZY: If you are talking about 23 deficiencies in the way a utility is being managed or how 24 the programs which are operating the reactor are being l 25 managed, I think that is outside the scope of this
- 0
_ _ _ _ _ _ _____-__________.___________________________.___.._______.-____m___.-__._
117 1 program.
2 Relative to your other comment that the staff 3 does not have a copy of the IDCOR methodology and it is 4 already being applied to plants, the application is part 5 of the development of the methodology. This is not the 6 application what will be done after we have had an 7 approved or accepted methodology. The best way to develop 8 a methodology is to try actually how would it work to get 9 feedback from the utilities.
10 It is not IDCOR who is applying it to the 11 utilities. The utilities themselves are applying it.
12 They are feeding back information wherever they have
,e7 13 difficulties and where it is clear to them what the i
14 methodology has them to do. That is being factored into 15 the revision of it and only after they factor this in do 16 they feel that they have now a methodology.
17 MR. CARBON: Question. IDCOR has done four or 18 five PRAs, and I would guess that it may very well in this 19 list of questions you speak of ask questions or point out 20 questions to the utilities based on the results that they 21 found in the four or five PRAs that they have already 22 done.
23 Do you have any confidence, or do you know yet, 24 are they really going to look for outliers and are they 25 going to be effective in finding outliers with this 8
l 1
118 sJ I methodology?
2 MR. ROSZTOCZY: IDCOR has done detailed analysis 3
for four plants. They really did not do PRAs. They 4
simply picked up four plants for which a PRA was 5
available. The PRA could be the WASH-1400 PRA or it could 6 be something more recent than that. So they analyzed 7
those four plants in detail. After that they developed 8 this methodology, and this methodology is supposed to have 9
now owners of other plants to search for vulnerabilities.
30 Obviously, on those four plants where the II detailed analysis is already available, and now this 12 simplified methodology is being applied, you don't expect 13 to learn a lot in terms of vulnerabilities. You are
- J
' ~ '
14 learning more along the lines of whether the methodology 15 is working and whether it is easy to execute it.
16 However, there are three other plants where 17 there is no PRA available and this methodology is being la applied. So what type of results they get on those three I? plants that could be used as part of their evaluation in 20 terms of whether the methodology lo able to pick up 21 vulnerabilities, up to now from discussions what we had 22 with them, they are rather confident that this methodology 23 the way how it is being proposed will be able to pick up 24 most of the outliers that the PRA would pick up.
25 MR. CARBON: Are they going to do PRAs in these 8
119
(
I other three plants?
2 MR. ROSZTOCZY: No.
3 MR. CARBON: Do you share their confidence that 4 they are going to be able to pick up outliers?
5 MR. ROSZTOCZY: We will have to receive the 6 details of the methodology and review them, and by next 7 fall we can give you a very good evaluation.
8 MR. OKRENT: Are you committed to using the 9 IDCOR methodology?
10 MR. ROSZTOCZY: No. The idea is that at the end 13 when we have finished with this, then there would be a 12 generic letter issued to each plant owner. The generic
-) 13 letter would specify what is expected from it in terms of 14 further search for vulnerabilities in the severe accident 15 area. It might be different for some utilities than for 16 others. For example, it would be different for someone 17 who already has a PRA and different for someone who does 18 not have a PRA.
19 This generic letter then would also mention what 20 is an appropriate way to perform this examination, and 21 what we hope to say is that the IDCOR methodology with the 22 appropriate changes, whatever evolves in the six months' 23 review what we are planning, would be an appropriate way 24 to do this evaluation and we hope that we will be able to 25 Adentify an acceptable way to do the examination.
8
120
<-m
- I MR. OKRENT
- Let me ask one more question. In 2 reading the draft document, which is dated February, it 3
says that you expect to get some reports in March.
4 MR. ROSZTOCZY: That 10 correct.
5 MR. OKRENT: Presumably this will also include 6 the methodology that was used in preparing the report; is 7
that correct?
8 MR. ROSZTOCZY: Yes.
9 MR. OKRENT: So in March maybe we could also 10 have the benefit of these reports on methodology?
II MR. ROSZTOCZY: That is almost correct. That 12 was the schedule which we put down in January when the
,-w;
, 13 document or our plan was prepared. It turns out that we
( ^ ^')
14 just had a meeting with IDCOR carlier this week and they 15 indicated that there are a few weeks delayed. So probably 16 they will be coming in in April, 17 MR. OKRENT: Thank you.
18 MR. EBERSOLE: Zoltan, let me ask a question.
19 Back at the program element page you had a statement under 20 1, development of guidelines and criteria for plant 21 examination, and then you subsequently mentioned 22 walkdowns. Well, I hope you develop a manual or a 23 handbook of what you look for when you do a walkdown, 24 because a walkdown is worthless unless you first know the 25 plant in very high regard indeed as to what is behind the 8
121 s
ss I walls. You don't know that the battery room is on the 2 other side of the boiler room or anything because ---
3 MR. KERR: You realize that the walkdown is 4 going to be done by the plant people. This is not going 5 to be done by an outside group.
6 MR. EBERSOLE: The plant people that I know know 7 how to run the plant and push the buttons, but they don't 8 know how it went together.
9 MR. KERR: But I mean this is going to be done 10 by employees of the utilities. It is not going to be done Il by an outside group.
12 MR. EBERSOLE: Those are precisely the ones that a
g~y 13 I tell you don't know where the cables are or where the
(
14 boiler room is relative to the battery room or what is 15 behind the brick wall. They don't have a perspective view 16 in this safety context that you are talking about. So it 17 is imperative that you develop a manual for walkdown 18 procedures with due regard to prior in-depth understanding 19 of the multi-feature aspects of the plant which never show 20 on drawings.
21 MR. ROSZTOCZY: We share your views in the 22 walkdown, and we expect that the IDCOR methodology will 23 have in it some procedures for the walkdown and we intend 24 to look at that.
25 (Slide.)
8
~
122 m.--
q
_U 1 The next slide shows in gross terms the schedule 2
that we set forth on which we would like to accomplish our 3
goals, d
The date shown is the accomplishment of the 5
reference plant analysis. NRC is presently analyzing six 6
reference plants and we expect that most of the analyses 7
will be completed by June of this year.
8 In addition to analyzing the six plants, 9
analyzing severe accident sequences for the six plants, we 10 are also performing a sensitivity analysis or uncertainty Il analysis for these plants and that is supposed to be 12 finished by July.
,7 q 13 When we received the IDCOR evaluation of the i" ~ " "/
14 four IDCOR reference plants, we have provided some 15 comments and some feedbacks to IDCOR, and they identified 16 a set of technical issues which were the most important to 17 the final conclusions and those where the differences were 18 the largest between IDCOR's thinking and oura, we lifted 19 these out as technical issues and we working on them and {
20 {
expect to arrive to some kind of a resolution on these by 21 July also.
22 The IDCOR methodology is expected to be 23 completed by October, and then we expect to come up with a 24 proposa3 in terms a generic letter and this would be 25 communicated to the utilities and brief the Commission on 8
123 s-
' I that by the end of the year.
2 In terms of the future plants, the guidance, 3 what we mentioned earlier on the role of the PRAs, that 4 would be prepared and issued for public comment by 5 February.
6 If there are any other generic changes that 7 emerge during our development of the guidelines and 8 criteria that need immediate attention, then we would 9 issue those for public comments a couple of months after 10 the guidance on the role of the PRA. So probably in 31 April.
12 This schedule does not show the completion
,c s 13 schedule for the source term related rule changes. I will J
'~'
14 come back to that later and I will show what is our 15 schedule on that.
16 (Slide.)
17 As I mentioned, the plan is basically composed 18 of three elements. For each of the elements we have kind 19 of an outline of the major parts of that element, the 1
20 major tasks which feed that element.
21 The notation what we used here is you see two 22 kinds of symbols, solid squares and dotted ovals. The l
23 solid squares represent a certain subtask of this 24 program. So everything what is in solid squares is part 25 of the work that is being done under this program 8
124 J I element.
2 The ovals represent input to the program from 3
somewhere else, for example, from the research program or d
from the industry or from IDCOR, but it is information 5
coming from somewhere else into this program and these are 6 the elements of the program.
7 If you start back from the right-hand side of 8 the screen, then our goal is to arrive to this generic 9
matter and the Commission briefing on that by the end of 10 the year.
II The two main components is supposed to be the 12 review of the IDCOR methodology and development of the ry 13 guidelines and criteria by the staff. Obviously there is
(" ~ ' "!
Id some connection between the development of these two.
15 Let me discuss first the IDCOR part. The IDCOR 16 methodology is supposed to come into us starting in April 17 and finishing in July, and we are right now preparing 38 instructions for the review of the methodology. So these 19 two together should be ready in about April and then we 20 will start the review itself and we hope to. finish it by 21 10/86.
22 In terms of the guidelines and criteria, 23 strawman guidelines, we already started to work on the 24 preparation of strawman guidelines. This will be prepared 25 separately for each of the six reference plant types.
8
125 1 Those six plants were selected to represent different 2 designs, especially different containment designs and then 3 the expectation is that the actual plants can use and can 4 rely on the reference plant review as much as possible.
5 The last of this is supposed to be finished by 9/86, and 6 some of the earlier ones will be available earlier.
7 Proposed criteria will be developed kind of 8 together with the guidelines. Following shortly after the 9 guidelines we also expect to finish that by September, 10 The evaluation of the reference plants of course drives 11 this, and that should be available this summer, and the 12 reference plant analyses and sensitivity analyses coming
~
13 from Research feeds into that. Resolution of the
(' ' ';
14 technical issues is also very important.
15 (Slide.)
16 In our implementation plan we also provide a 17 more detailed schedule that indicates not only the end 18 dates of each of these, but some of the earlier dates that 19 we have to meet in order to get to those end dates.
20 The standards for the acceptable methodology, 21 for example, are being developed this month, and the IDCOR 22 submittal was expected originally in March, which we will
?3 be a few weeks later now.
24 We expect to provide comments back to IDCOR on 25 the methodology this summer and then finish the review.
8 l
1 l
126
_> 1 The schedule is broken down to individual plants 2
and some of the plants which will be available earlier 3
will be reviewed earlier and the ones coming later will d
follow.
5 MR. WARD: What do you mean by the uncertainty 6
ranges? Can you define the uncertainty ranges?
7 MR. ROSZTOCZY: I am sorry. I am not sure which 8
one are you referring to?
9 MR. WARD: Under technical issue resolutions.
10 MR. KERR: 3.2.1.
Il MR. ROOZTOCZY: The technical issue resolutions?
12 MR. WARD: Define uncertainty ranges.
13 r- MR. ROSZTOCZY: Oh, define the uncertainty
("~ )
14 ranges for the reference plants. We are in the process of 15 identifying those inputs to the calculations which we 16 believe are resulting into the largest uncertainties to 17 the final results. The final results are the core damage 18 frequency release from the containment and risk.
19 After identifying each of the most parameters, 20 we have to establish a range for each of these what we 21 consider a reasonable range of what the parameter could be 22 in the real world, and then we combine together the 23 uncertainties from these in terms of the final uncertainty 24 on the results.
25 So these uncertainty ranges which will be input 8
( >
127
(
-/ 1 to the sensitivity calculations or uncertainty 2 calculations has to be identified for each of selected 3 parameters and that is presently going on. It has been 4 going on for two plants and now it is going on for a third 5 plant.
6 MR. WARD: There is a large body of operating 7 experience for let's say Surry that has been running for a 8 good many years. Grand Gulf hasn't been and there would 9 be a much smaller body of operating experience. Does that 10 mean you end up with a smaller uncertainty range for Surry 11 than there is for Grand Gulf?
12 MR. ROSZTOCZY: No. When, for example, we
,, ~ . 13 consider an uncertainty range for an operator action, then t :
14 we are locking at all available data which would be useful 15 to that operator action, data coming from nuclear planta, 16 data coming from other sources, from non-nuclear plants 17 and it altogether helps us to decide what is the 18 appropriate part.
19 The part where the individual plant features 20 come into play is that our people do go out to the plants, 21 to the reference plants for a walkdown and they actually 22 inspect or see what does the operator have to do to 23 perform this function.
24 MR. WARD: If the data for operator actions 25 comes from all plants, non-nuclear plants and everything, 8 .
128 L_J l you certainly aren't identifying outliers in the 2 performance of operators at a given plant. ,
l 3 MR. ROSZTOCZY: We will be identifying outliers d
in the form, for example, that if we go out to Grand Gulf 5 and we look at what does it take to accomplish venting the 6 containment, and if we find that the operator has to be 7
sent far out somewhere to do five or six different actions 8 and there is no time to accomplish those, then that would 9
be a plant specific finding for that plant, or if we find to that this cannot be accomplished because the space, the 11 room that he is supposed to go in at that time is just not 12 available, that the hydrogens burns in that room at the
<- 13 time, then we would point out that that is not ---
i l
~* 14 MR. KERR: Let me point out that what he is 15 trying to define is a range of uncertainties, Dave, and 16 for that I think you do need to look at more than one 17 plant in a generic sense.
18 MR. WARD: Okay.
19 MR. EBERSOLE: Zoltan, I heard you say that in 20 the context of believing that perhaps that the operator, 21 you know, like Horatio at the bridge, will always save the 22 day, but there may be design features, and in particular 23 you have got Sequoyah up there, a plant which even the 24 owner operators can't yet figure out whether it is safe to 25 run or not. What are you going to do with their lack of 8
129
' I confidence in their own plant evidently hidden behind some 2 curtain that they don't know how to Ijft?
3 MR. ROSZTOCZY: Dr. Ebersole, I am not sure if I 4 follow the whole question, but what we intend to do is to 5 identify those operator actions what we think are 6 essential for safe handling of severe accidents. If once 7 these are identified, then the utilities will be asked to 8 change their emergency procedures in such a way that the 9 information needed for the operator to do this would be 10 included in the emergency procedures. Obviously we are 11 asking them to check if those functions can be 12 accomplished, g, 13 We expect that in a number of cases we will find i s
' 14 that it cannot be accomplished with the way how the plant 15 is now and some changes will be needed in order for those 16 functions to be accompliehable.
17 MR. EBERSOLE: Well, you are going to start with 18 a plant in a bad state in the first place.
19 MR. KERR: They aren't going to try to evaluate 20 a plant that has been shut down because the equipment is 21 not environmentally qualified.
22 MR. EBERSOLE: Your point of beginning is to 23 arbitrarily say that the plant has gotten itself in a bad 24 state in the first place.
25 MR. ROSZTOCZY: No. Our starting point is that 8
130 t_J l the plant was designed to handle obviously normal 2 operation transients and a certain set of accidents which 3 we call postulated accidents under the design basis 4
requirements. That is what it was designed for.
5 Now what we are looking for is we are asking the 6 question are there any other accidents more severe than 7 those that have a reasonably high probability of 8 occurrence so one should pay some attention to them in 9
addition to the design basis accidents, and we expect to 10 identify selected sequences for each plant type which Il falls into this category.
12 Once we identify those sequences, then we will q
13 ask the question, you know, what should the operator know 14 about these sequences and what should be in his training 15 program and what should be in the emergency procedures 16 that he is going to consult in case of something 17 happening, and our expectation is that we will find a la number of things which are presently not there but should 19 be there just for the purpose of the severe accidents.
20 MR. EBERSOLE: Okay. Thank you.
21 (Slide.)
22 MR. ROSZTOCZY: The second element of the 23 program is the development of the guidance of the PRAs for 24 the future plants. On this one our goal is to have this 25 approved by the Commission and issued for public comment 8
131
> 1 next February. So it will be less than a year from now.
2 In order to accomplish that, we hope to send our finding 3 up to the Commission about a month earlier.
4 It will basically address three parts, the 5 deterministic requirements which already exist, but they 6 touch on severe accidents, the guidance on the minimum 7 content of how much information needs to be in the PRAs 8 that future plants have to prepare and criteria for the 9 regulatory use of the PRAs to tell them how are we going 10 to measure the acceptability of the PRAs.
Il The deterministic part is basically coming from 12 available requirements, requirements wh'ich have been 7 13 established on USI's, unresolved safety issues and generic
'# 14 safety issues, and also some experience that has been 15 gained in the review of some recent PRAs like GESSAR and 16 the Indian Point case. Based on this we will establish 17 those requirements.
18 The guidance for the minimum content is going to 19 rely on procedures for core damage frequency and those 20 procedures for containment analysis and consequence j 21 analysis.
22 We have various guides available to date in 23 terms of probabilistic safety assessment and the IREP 28 procedures guide, the IDCOR methodology we will have also 25 and we intend to use it, and there are also guidelines and
132 r 1 I
2 criteria that will be available on the individual plant 2
examinations.
3 In terms of the criteria, we obviously are going d
to use the safety goal, whatever form the Commission 5
decides to issue it in final form, and we are also going 6 to use any containment performance objectives, if such is 7 available and in whatever form it is available. The 8 inside reports available will be also considered for the 1
9 development of the guidance.
10 MR. MOELLER: When you say content of the PRA, Il you mean the numbers that result or what went into it or 12 the methodology?
r ,, 13 MR. ROSZTOCZY: One can perform a PRA and can
! I
' 14 perform it just kind of let's say a cheap PRA, or you can 15 do a much more detailed PRA, or you can do a PRA when you 16 go into great detail in some areas and you kind of short-17 cut other areas.
18 MR. MOELLER: And that is what you will be 19 doing.
20 MR. ROSZTOCZY: That is what we are looking 21 for. We are looking at best experience with PRAs. We 22 tried to find out which areas were really useful and which 23 ones are the ones that don't contribute too much to the 24 safety problem, and based on that we want to specify how 25 much detail they need to go into in the areas which are 8
133 1 important and hopefully it will provide sufficient 2 guidance that the PRAs coming in in the future will be 3 acceptable.
4 MR. MOELLER: Thank you.
5 (Slide.)
6 The third program element is the changes in 7 rules and regulatory practices. A portion of it, what we 8 are already working on, is source term related rule 9 changes. The source term as such enters into our rules 10 and regulations at many different areas. We have already 11 reviewed those and we are right now selecting out some of 12 them what we consider more urgent or some that we have the
,e 13 information available on so we can initiate work on them,
- 4 14 and we already started to work on those.
15 In order to end up with source term related 16 changes, we have to replaced the old source term, what we 17 had, with some kind of new source terms. Part of our 18 program is the development of new source terms. I 19 Before we could develop these new source terms 20 we have to have some general approach, some general 21 principle that we are going to use for the source term 22 related changes, and then we can execute that across the 23 board in a systematic fashion. So we are working on a 24 regulatory principle along these lines.
25 Of course, we have to have the capability to do 8
134 1
t J I source term calculations. That has been accomplished, and 2
we are doing source term calculations.
3 The second area is the containment related d
changes. The containment is one of the main protective 5 functions in terms of the severe accident, and because of 6
this we are paying special attention to the containment.
7 We are asking the question of whether there is a 8 need for containment performance criteria. So far it 9
appears that the likely uncertainly will be a yes, and if 10 that is the answer, then we are going to develop the Il containment performance criteria.
12 MR. WARD: How are you going to decide that, r -, 13 Zoltan? You said you are asking the question and the r i
\" '
14 answer might be yes, but how will that be decided?
15 MR. ROSZTOCZY: The way how we are developing 16 the guidelines is we are looking at those sequences which 17 are the main contributors to the severe accident risk, and 18 then we are identifying the important functions which need 19 to be performed and the important systems which are 20 performing these functions, and are trying to establish a 21 yardstick, a measurement to say when are these acceptable.
22 By doing this and doing it systematically going 23 through, we will see whether the existing requirements 24 provide a sufficient yardstick to measure containment 25 related functions. If the answer would be that they do lif l
135
_/ 1 provide a sufficient yardstick, then there would be no 2 need for a criteria, and then we would come up with a 3 conclusion that it is needed.
4 If we find that it is very difficult to decide 5 on some of the containment related functions whether it is 6 acceptable or not based on present regulatory practice, 7 then we would recommend to up the present regulatory in 8 terms of some kind of a criteria. Basically I think what 9 we find about the criteria, what we need to answer these, 10 to respond to the guidelines, those would be the basis for 11 the containment performance criteria.
12 MR. WARD: Could you give me an example of one y- 13 or two existing, whatever you call them, that you might
~
14 use at present to judge the containment performance?
15 MR. ROSZTOCZY: Take containment sprays, for 16 example. If we end up with three or four sequences for a 17 given plant which are predominant, the most important in 18 the severe accident space, then we would ask the question 19 now do we rely on containment spray for these sequences.
20 And let's say we find two of them where we heavily on 21 containment sprays. Then we would go on and ask the 22 question how long do we need these containment sprays and 23 what capacity these containment sprays ought have.
24 If it turns out that the current existing 25 requirements provide sufficient guidance there, then we 8
136
~l
~;
I just put that there. We would say the utility ought to 2
check whether they have a containment spray that is going 3
to function during this sequence, and they ought to check 4
whether it is meeting these requirements, what they pull 5 out from the existing practice.
6 If the existing practice does not provide 7
guidance, then we will try to tie it down to the guidance 8 for those. Then when we put together these containment 9
related guidances into a group, then that will be the 10 basis for establishing some containment criteria.
II MR. SHEWMON: To continue that, leak rate would 12 be another one where presumably you have criteria, but
,r- 13 whether or not it would cope with something beyond the
- "qi 14 design basis for pressurization would not be?
15 MR. ROSZTOCZY: That is correct. There is a 16 leak rate requirement. So when you arrive at some 17 question, which really is the leak rate, then you would 18 look at this leak rate requirement and you can arrive to a 19 conclusion that this does the job. You can arrive to a 20 conclusion that it doesn't do the job, or you can arrive 21 to the conclusion that it is too restrictive, and really 22 there is no need to restrict the leak rates to that much 23 because one can take ten times a' much leak and it is 24 still acceptable. There as still no real problem.
25 In that case we might even suggest to change the 8
137
- 1 existing requirement and relax some of the existing ones 2 since it doesn't have to be as tight as it was set.
3 MR. SHENMON: Okay.
4 MR. ROSZTOCZY: The last area is what we call 5 other severe accident related changes. These are the 6 changes what we don't know and what might come up from the 7 review, but if things come up from the review that need to 8 be incorporated into our regulatory practice, then we will 9 find the appropriate way to incorporate it at that time.
10 (Slide.)
II As I mentioned earlier, we looked at where does 12 the source term enter into our regulatory practice, and we f- 13 identify 12 separate areas. After looking at the 12, we
)
14 decided that two of them were the type where any change or 15 need for a change was not needed or the source term didn't 16 play a significant role in any change, we ended up with 10 17 which are the potential areas for change.
18 These 10 then have been grouped into three 19 categories.
20 The first category is the near-term. This 21 includes items where we are either already practicing 22 something that uses some new source term information, or 23 we feel that work can go on to prepare for some kind of a 24 change pulling together the information that is needed for 25 that change.
8
138 7
J I These three are the revised treatment of an l 2
accident in the environmental impact statements, the 3
removal of spray additives and credit for the suppression 4
pool scrubbing effect. The first one is already being 5
used and we have initiated work on the second and the 6 third.
7 The second group are the intermediate changes.
8 These are changes that we felt we need the reference plant 9
analyses. We need the evaluations for the six reference 10 plants to see what do we learn from that in terms of II source terms and the uncertainties associated with source 12 terms. As soon as that is available, then we will be gq 13 ready to proceed with these six.
- i. J
"~"
14 Then, finally, the long-term changes. Those 15 were the ones where we felt that there is no real 16 urgency. So they probably can be at the end of the list 17 and we will get to them as soon as we finish the others.
IB (Slide.)
19 I mentioned earlier that it is important that 20 when you do these changes, and especially if you do it 21 piecemeal, one by one, the changes as we intend to do it, 22 and it is important to keep in mind of what are you doing 23 overall and have some general principle that you are
[
24 following so these changes are going to complement one 25 another rather than contradict one another.
8
139
_/ 1 We have been looking at various approaches that 2 one could take to replace the present approach. The 3 present approach, what we are using in most parts of the 4 regulation, is a TID type of source term, which is an 5 arbitrary source term, combined with a design basis 6 accident analysis for the containment in terms of pressure 7 and temperature.
8 And then Regulatory Guides 1.3 and 1.4 provide 9 some information of how you can use this for inside to containment releases and how to use this for outside 11 containment releases. So the basic assumption is 12 acceptable under the present regulations.
,c y 13 The basic question is that based on the new
'~'
14 information available now are we better able to predict 15 what might happen in the plant. One would like to see a 16 more realistic approach to the source term than what was 17 originally developed.
18 The original approach, the TID approach, is 19 simply, which has great advantages, it is inconsistent and 20 it is arbitrary. But, nevertheless, it has done a good 21 job for many years. If we want to replace it, we should 22 be very careful in what are we going to replace it with 23 and we should think it over twice before we go to 24 something new.
25 MR. WARD: When you say it has done a good job 8
140 c- --
- l I for all these years, how do you know and why do you say 2
that?
3 MR. ROSZTOCZY: Because even based on today's d
knowledge, what is available to us, and some of the 5 calculations which are coming out now in general are 6
indicating that those requirements which were put forth 7
were sufficiently conservative. They did not create a 8 circumstance when maybe unsafe designs would have been 9
approved. There are a few new things, what we know today, 10 which would be added to certain designs like maybe some of Il the lodines should be emphasized as much as they were 12 emphasized back a few years ago, but overall they are not
,x 13 too far off. They may be a little bit more demanding in
! )
"^'" Id some areas and may be a little bit lenient in others, but 15 within the simplicity what we have there you cannot expect 16 to be exact. So altogether it is not bad.
17 As a matter of fact, it is our expectation that 18 even after we finish, we probably will be able to leave 19 this approach as an option, which would be a very 20 practical thing in that nobody would have to go back and 21 redo something because the old one is unacceptable. The 22 old one would stay acceptable and there would be new ones 23 added to it that they could use, which might be more 24
- favorable to them for certain purposes.
25 So we have looked at various approaches, some of 8
+ Q,* .
141
- I them more drastic and some of them basically on the old 2 one and just updating it.
3 (Slide.)
4 One what we are arriving at, that we are 5 favoring at the present time, is the one described on this 6 slide, which is basically following the same approach as 7 was done on that, the old TID and Reg. Guide 1.314 8 approach, but updates it to today's knowledge and 9 everywhere where it is possible to make it more realistic 10 than it has been in the past.
11 So instead of having arbitrary numbers that you 12 release, we would permit actual calculation of the r- 13 releases and use the actual calculated numbers with some 14 appropriate account of the uncertainties.
15 In addition to permitting a detailed 16 calculation, which is rather expensive to perform, we also 17 intend to develop a simplified source term based on the 1
18 available information and in looking at all the available l 19 source term information together and trying to bound it in 20 some fashion.
21 In this case then the simplified system would be 22 available to utilities as an option and if they wish they 23 could use it. Obviously when you come up with a ,
1 24 simplified one, there is some penalty, some price what you l l
25 pay for it. There will be some more conservatisms in the !
8
142
-J l simple one than in the detailed calculations, and that 2
would be the price what they would be paying for it.
3 We are also giving special attention to some of 4
the areas that require radioactive material releases 5
insido containment as opposed to releases from outside 6
containment. This would be the areas of equipment 7
qualification and also the design of filtering systems 8 that you basically base on what is the inside containment 9
content of the gases.
10 (Slide.)
11 We also have to have some kind of a limit that 12 kind of drives your requirement, like your leak rates from 13
-q the containment, and it would be our intent to keep it in
- i
"~~ 14 a similar form as presently exists in Part 100. There 15 might be some need to update the numbers or maybe to 16 change somewhat Part 100, but basically the same approach 17 would be followed.
18 (Slide.)
19 The next slide shows the schedule for the 20 various source term related changes.
21 MR. KERR: Excuse me, Zoltan. You are saying 22 then that the environment being considered in the 23 calculated leak rate would be a DBA environment and not a 24 severe accident environment?
25 MR. ROSZTOCZY: No, sir. To the contrary, it 8
143 l m/ 1 would be a severe accident environment. We would consider 2 a number of severe accidents and accident sequences 3 leading to core degradation. So severe accidents, select 4 all the severe accidents which are important and which are 5 driving the risk, and then calculate the source term for 6 these severe accidents, and this is*the one that would be 7 used for determining the leak rate.
8 MR. KERR: I am not talking now about the 9 fission product release in the containment, but rather the 10 temperatures and pressures that would accompany that which II would also determine the leak rate in an actual accident.
12 cw 13 MR. ROSZTOCZY: That would be done i ]
14 realistically. Maybe it is not completely clear, but the 15 second bullet is supposed to address part of that. We 16 would permit a realistic calculation of that for these 17 severe accident sequences.
18 MR. KERR: Okay. I understand. Thank you.
19 MR. MICHELSGN: Are you going to leave that 20 slide now?
21 MR. ROSZTOCZY: Yes.
22 MR. MICHELSON: Maybe I can ask my question. It 23 hasn't been covered and it is perhaps related here as much 24 as anywhere. This morning we discussed the GDC core 25 modification in terms of this new broad scope rule about 8
144
_J I pipe break. In the old accident scenarios or sequences 2
you used to consider, you considered double-ended ruptures 3 and uncertain pieces of equipment didn't work and that is d
how you got your severe accident numbers. These were all 5
beyond the design basis.
6 Now that we may change our design basis to that 7 of a leak and not a break, how does that affect this work 8 and how does it affect the probability, keeping in mind 9
that come equipment is being removed like pipe whip 10 restraints. So now if you do postulate the double-ended Il rupture, the probably of loss of containment might be a 12 great deal higher than it was before because there is r~y -
13 nothing to restrain the pipe from breaking right through
! )
~~
14 the containment or from the jet impinging on the wall and 15 damaging the containment and so forth.
16 Is that going to be picked up and accounted for 17 some time in the future?
18 MR. SHEWMON: Well, if you do, be sure you get 19 the probability of the guillotine break right.
20 MR. MICHELSON: Well, I assume it hasn't changed 21 any, but perhaps ---
l 22 MR. SHEWMON: The probability hasn't changed?
23 You are probably right, it hasn't.
24 MR. MICHELSON: But perhaps removing the l l
25 restraints has now decreased the probability of the l
Ill
145
't I failure, but it has increased the probability that the 2 break will cause certain kinds of damage because it is 3 beyond the design basis, whereas before today it is not 4 beyond the design basis because we restrained the pipe.
5 MR. KERR: Carl, I don't understand your 6 question. Maybe Zoltan did. One would calculate an 7 accident sequence, which would include, for example, a 8 large pipe break. So in that sense if they chose that 9 sequence, presumably one would get the appropriate 10 temperatures and pressures which would be characteristic il of particular plant design. Did you question beyond that?
12 MR. MICHELSON: Yes. What the present
<x 13 calculations do is they put in a probability that having
- i. }
14 experienced such a break that you will also lose the 15 containment, and that probability has now changed because 16 you ---
17 MR. KERR: But here if you were calculating la containment performance you would get that probability I 19 assume by taking the environment that would exist in the 20 containment as a result of that sequence and then looking 21 at containment performance.
22 MR. MICHELSON: Well, they look at things like 23 isolation valves failing and ---
24 MR. KERR: All of the things that would enter 25 into containment performance. Now is your question beyond 1
146 tJ 1 that in some fashion?
2 MR. MICHELSON: Yes, it is because presently the 3
design basis is to restrain pipes from puncturing the d
containment. The new design basic will be not to restrain 5
pipes.
6 MR. KERR: But one would have to do the PRA, 7
depending on what is in the plant, I assume, wouldn't one?
8 MR. MICHELSON: The question was are you going 9
to go back and -- well, is this program going to do that 10 PRA or are we going to do another set or how is it going Il to be considered?
12 MR. ROSZTOCZY: This program basically evaluates 13
,e six plants, six reference plants.
~~
14 MR. MICHELSON: As they stand today.
15 MR. ROSZTOCZY: As they stand today, or it gives 16 me credit for changes which they have already committed 17 to. Like under the ATHS rule they are planning to put 18 something in in a year and a half from now.
19 MR. MICHELSON: My question is related to if 20 they adopt the broad scope rule and if the plants go in 21 and remove the restraints and so forth, how are you going 22 to evaluate the effect on this work?
23 MR. ROSZTOCZY: I can give you some ideas how it 24 can be done, but I don't believe that it is included in 25 the work right now.
8
147 1 Right now the plant are evaluated as it stands.
2 If that plant a year from now makes a decision to take 3 this out, then what would be available from here is there 4 will be available some calculations for the risk of what 5 happens to the containment as it is.
6 If at the time when they consider taking it out, 7 they revise the numbers of what is the risk, and one could 8 feed the revised numbers in this analysis and see what 9 does it do to the end results. You can read out either 10 that is a very important item to the end results or you Il might read out that that changes one of the sequences by a 12 factor of three, but that really contributes ver*y little f-\ 13 to the overall and therefore it is not important. You i !
provide the means to feed it in, but right now it is not 14 15 being done both ways.
16 MR. MICHELSON: The point is that for the 17 present plant it is not even in the sequence because it is 18 not beyond the design basis. The design basis is to 19 restrain the pipes. So that kind of a sequence doesn't 20 exist.
21 MR. ROSZTOCZY: Right.
22 MR. MICHELSON: It will exist though if you take 23 the restraints out and you could go back and do the 24 arithmetic, but it is not apparently a part of this 25 program to do that.
8
148 l l
l t_l I CR. OKFENT: By the way, I might note in passing 2
that there was 6 report I think also done by Livermore, 3
but I am not quite sure, that looked at what they called 4
secondary caun,es of pipe failure, meaning the crane might 5
fall and hit a pipe and rupture it, or a steam generator 6
support might be inadequate and might topple over and 7 And those probabilities or cause a pipe break.
8 frequencies, if I recall correctly, were said to range 9
from 10 to the minus 5 to 10 to the minus 7 over the 10 plants that they looked at.
Il MR. SHEWMON: This was a steam generator coming 12 down and taking a pipe out, for example?
ry 13 MR. OKRENT: The indirect causes, and one end of
!. I
"~" 14 the range doesn't sound so bad, and one end of the range 15 falls within the kinds of sequences that Zoltan has said 16 they would be considering, and it is not clear what the 17 containment condition would be for those, whether it 18 depends on pipe restraints or other things.
19 MR. EBERSOLE: Zoltan, in the course of 20 developing the sequence of the severe accident I presume 21 you will take the, what I would think is the normal base, 22 which is at all times the operator will be trying to pour 23 Water on the core no matter what, is that correct, and he 24 may or may not succeed.
25 MR. ROSZTOCZY: I am sorry. I did not hear part 8
149 J l of your question.
2 MR. EBERSOLE: I am saying will in developing 3 the sequence of the severe accident you will take what I 4 think the operator would naturally do, which is to try to 5 always pour water on the core and never stop.
6 MR. ROSZTOCZY: That is a very important issue.
7 As it turns out on the plant, what we are working on now, 8 which is a BWR, Mark I, that is the case for all the 9 sequences but one, a response sequence when he has the 10 instruction to do the opposite, and we are looking very 11 carefully of what does this mean and what is the 12 likelihood that he will succeed in that one case, the one 13 which is different than his normal instructions. The
\ '
l 14 human factors people are looking at that in detail, and 15 the success of performing that function is a different 16 number than the success of the others.
17 MR. EBERSOLE: Okay, just so you are looking at 18 it.
19 MR. ROSZTOCZY: Yes.
20 (Slide.)
21 This slide shows the schedule for the completion 22 of the source term related changes. As you see, it is a 23 very tight schedule, and the two changes which I 24 mentioned, they are in the short-range program, we expect 25 to have something out for probably comment by September.
8 3
150 L- I-It turns out.that neither of those require rule 2
. changes. So those won't be rule changes. They are 3
changes in regulatory guides or standard review plans, but d
we intend to prepare the proposed changes by September.
5 The second ground, which was the intermediate 6
group that has a spread, some of those, especially the 7
ones which have a higher urgency, like the emergency 8
planning, we expect to finish by early next year and
(
9 issued for public comment'and others maybe by the summer 10 of next year.
11 Then, finally, the long-term ones, will be 12 toward the end of 1987.
rx 13 l
(Slide.)
k' 14 l This kind of completes the: description of the 15 three basic elements of the program, and now I would to 16 say a few words about the relationship between our 17 program, the implementation program, the severe accident 18 implementation program and other programs.
19 The first area that we are interfacing with and 20 depend rather heavily on is the research program. Most of 21 the severe ~ accident data and information that we have 22 today has been developed in NRC's reserach' program, severe 23 accident research program.
24 The research program itself is described in 25 NUREG 0900, and a supplement to that which will update the 8
4 151
\ -
1 plans, the research plans for the coming two years will be 2 issued shortly. That is the NUREG 0900 supplement. I i 3 believe it.is scheduled to be issued in April.
l 4 Another. report, NUREG 0956, has been issued on 5 -the. technical ba' sis of estimating source terms. This is 6 out for public comment. A large number'of public comments i
7 have been received, and it is going to be reissued this 8 summer.
1 9 Then, finally, the third report that is going to 10 document the reference plant analyses for the six il reference plants and the uncertainty study that is 12 associated with that is NUREG 1150. This one is scheduled f
13 to be issues probably in early fall of this year.
14 In terms of NRR programs ---
- 15 MR. WARD
- Zoltan, I have a question on the 16 research program. Earlier in this meeting, today and 17 yesterday, we discussed the safety goal and the core melt 18 quantitative goal which might be a part of that, and one 19 of the big problems with that or with implementing that or i
20 understanding what it means, is with the definition of 21 core melt where some people say that the definition that 22 is wanted as far as the safety goal is concerned is a core 23 melt that goes through the vessel is in effect a challenge 24 to the containment, whereas the core melts that are 25 calculated in PRAs don't seem to be that, but it is a l
....-,-~..~,,,,--._-,-,,.-_.,-.,,._,-_,_.,.-t.._. m,...,.-,,.-m-,-- _ - . - - , -
152 I
little uncertain as to exactly what they are.
2 Now there are questions raised about whether the 3
severe accident research that has gone on should have been d
ab13 to provide, you know, some sort of method for 5
sharpening the definition of core melt so you would be 6 able to calculate or know what you were talking about over 7
this spectrum from onset of loss of core cooling to 8 complete melt.
9 Is there any element in the research program on 10 severe accidents which is addressing that and is going to 11 provide that sort of tool, or do you see a need for that 12 sort of tool?
cy 13 MR. ROSZTOCZy: Jocelyn is here from Research 1
" 14 and in a moment we will get her understanding, but let me 15 give you my understanding.
16 The basic question is whether a core melt 17 represents just damage of the core or does it represent 18 something when you also melt through the vessel.
19 At the present time the PRAs and the analyses, 20 what we are getting, do not separate between these two.
21 They kind of group them all together and we intend to use 22 it along those lines.
23 It may be of some benefit in the future if they 24 would be looked at separately. Probably there would be 25 some benefits from it, but it is not something that is 8
4 153
/^[s
(_ l essential to our-work. It might be more important for 2 other work like, for example, to decide what number you 3 put into a safety goal.
4 MR. SHEWMON: What you are saying is that once 5 the fuel becomes overheated or we have once lost the 6 ability to coo the fuel that we conclude what, that it 7 runs out on the floor inevitably and challenges the 8 containment or that it goes through the containment or 9 what is inevitable once we have this overheated fuel?
10 MR. ROSZTOCZY: At the present time we are II looking at partial results from the evaluation of severe 12 accidents and then the final result.
13 The first partial result is core melt
'"' 14 frequency. You calculate a number which is the calculated 15 core mel*t frequency.
16 MR. SHEWMON: And that means that we can no 17 longer cool the core or what does that mean?
18 MR. ROS7TOCZY: In the context how we use it it 19 means all of this. It means that you have exceeded the 20 core cooling limits and therefore the core is damaged, but 21 whether it is damaged as much as Three Mile Island, and 22 that is maybe damaged less, or it melted through the 23 vessel it not separated out for the purpose of getting the 24 core melt frequency. All of those are added together and 25 that is she core melt frequency.
8
-,_,,.,,-_._7 -- _, .y., -_- .,%. , . , . , - _ . _ _ . , _ _ _ , _
l 154 r,
_) 1 We think for our purpose we can work with it 2 along those lines. Then from there we go on and see what 3 more will happen.
4 MR. SHEWMON: I understand one of the 5 differences between what the NRC has as a position and 6 IDCOR is what reactor operator behavior or response one 7 should expect under these conditions. Is that anything 8 that you are considering in your program?
9 MR. ROSZTOCZY: Yes. As a matter of fact, we 10 are working with them on some of those and we have asked 11 them to show the documentation and the basis of how do 12 they arrive at the numbers when they give credit for 13 operator action, and our people are also doing the same.
, "i 14 I think there is some improvement in that area, but I 15 expect that we will see some differences at the end, too.
16 In the sensitivity studies we account for this and we show 17 how does it affect the result.
18 Jocelyn, would you be interested to say anything 19 more about it?
20 MS. MITCHELL: Not unless there is a question.
l 21 MR. WARD: I have a question. Is there a part 22 of the severe accident research program which is directed 23 toward developing or providing a tool to allow PRA or an 24 accident analysis to discriminate among these various 25 classes of core melt?
8
. . . - . = ~ . . . . ~ . ~.
155 l (("h (,) . 1 MS. MITCHELL: I don't know. ' I am afraid I 2 don't know the answer to your question. That kind of-3 research would be in the Division of. Risk up on Nicholson t
4 . Lane and it is not something the fuels people are'doing.
I 5 Whether they are or not, I-don't know. I don't know the i 6 answer to the question.
7 MR. SHEWMON: Well, but if it was going on you 4
} 8 should have known~about it, or would know about it, i
9 MR. ROSZTOCZY: With all fairness to Jocelyn, 1
10 she is here basically to cover the source term area and
- 11 this is outside of her area. But why don't I check on i
12 this and we will let you know, Mr. Ward. We will find the 13 answer to that question.
O - $. 14 MR. SHEWMON: In all' fairness to Jocelyn, she is t 15 pretty good in what'is going on and how they do it, and if 4
16 any probabilist was doing it, they probably should be i
17 talking with people who model core behavior. But by all j- 18 means check.
)4 19 MS. MITCHELL: Thank you for the compliment, but i'
20 I can't accept it. I really don'.t pretend to be up on 4
1 21 everything, and I. don't want to say that nothing is going 22 on.- It is-just that I don't know about it, and I don't 4 23 know the answer.-
3 24 MR.:KERR: It-is my recollection, though, that i
- 25. over the years we have heard of the need-for experimental Ill 1
i 4
y,- -,,r- -
p.,.,,py 4. , .w..wwy.-m-r-+w. nm_-- ,,,w,,.ve m,,,m,,, .,_,-,r---.,.,--- , , , , mm,,,-ww.,,w,-..,-*.g..c., -e,,w,,#we.t.ww,w-s-,..-- r wo - e-
4 156 r~q k._j 1 'and analytical work on the behavior of fuel as it melts or 2 as it runs down channels and as it goes through the
, 3 bottoms of retaining crates. Hence, there certainly has 4 been both a great deal of expenditure of effort in both 5 the analytical and experimental areas. Now perhaps it has 6 not produced information or has not had as its objective 7 answering the question you raised.
8 I can only say that certainly one has gone 9 beyond the initial damage of the core in the research 1
10 program in what has been called progression of core il damage.
12 MR. WOODS: Jocelyn, your answer is sort of
, , 13 shifting it off to the risk analysts. I don't think they (4
"~
14 know that much about it. I mean I see that any answer to 15 that sort of question has got to have some sort of a 16 mechanistic or phenomenological component to it, and I 17 don't think the risk people know anything about that, do 18 they?
19 MR. ROSZTOCZY: I am sure this is not being 20 overlooked. It turns out that the Division Director for 21 that Divisdon is Mr. Ernst. He is also one of the key
^:-
22 members on the Safety Goal Committee. So he is keenly 23 aware of this question. He has been struggling with it 24 for the past year or so, and he is the one who directs 4
25 that part of the research program which would address 8
. - -_ _ _ . . _ . _ . . . _ - . _ _a _ _ _ . _ . - , _ _ . .. _ _ , - _ _ . . - . .
l 157 I this. He has the entire PRA program under him and he also 2 has under him all of the reference plant analyses what we 3 were talking about here.
4 So all of those are really in one hand. The 5 only thing what we present here do not know is whether he 6 found it useful to initiate such a program, and if he 7 found it a high enough priority or something that would 8 have a high enough return to justify it under current 9 budget circumstances. I think we will have to check it 10 and find out. The answer should be very simple to use.
11 We just have to ask him.
12 MR. WOODS: Okay. Well, we would appreciate m 13 that.
- MR. ROSZTOCZY:
14 So we finished the relationship 15 with the research program.
16 The second area is other NRR or NRC prograns.
17 Obviously the safety goal is one. We are looking forward
!8 to the Commission issuing the safety goal in some form, 19 and whatever form is issued we are going to use it in our 20 programs.
21 There are also a number of unresolved safety l 22 issues and generic safety issues which touched upon severe 23 accident, and we expect to use the resolution of those to l 24 the full extent in our work.
25 And, finally, we the PRA reviews, there have <
lli l
158 I}
(,,; I been many PRAs performed by now, over 20 PRAs for various 2 plants, and various things have been learned from that and 3 those insights are being collected.together and we expect 4 to review those.
5 The last part of the slide indicates the
- 6 relationship with the industry. We intend to conduct each 7 element in our program with some participation from the 8 industry.
9 In terms of the existing plant, this 10 participation is being done through the IDCOR program. We 11 have been working with them for a number of years now and 12 we certainly intend to complete that work working together
- 13 with them.
" 14 In terms of the source term related changes, we 15 have establish liaison with an industry group, an ad hoc 16 committee for this purpose, and how we are working very 17 closely with them in a somewhat similar fashion as we did 18 it with IDCOR on the existing plants.
19 In terms of the new plants, the role of the PRA 20 for the new plants, we have not established a contact with 21 the industry yet, but we are in the process and we hope to 22 set up something-for that also.
23 That completes the description of the 24 implementation program as it stands today. In the draft
- 25 version, what you have received, also there is a Ill
159 I discussion of external events and the question of how 2 external events should be considered.
3 (Slide.)
4 This was one of our major areas for the last few 5 months to consider and see what would be the best way to 6 include external events.
7 We are now at the stage when it has been a discussed on the office director level and there'is a-9 decision at least on that level of how we might proceed.
10 It is presently being documented and will be sent up to 11 the Commission in terms of a Commission paper in the near 12 future.
,. 13 The present recommendation of how to proceed is i
14 somewhat different than what you have received in the 15 handout. So today's slides, what you have a copy of, is 16 the updated version and I would like to bring it to your 17 attention that it is somewhat different than what is in la the written program.
19 (Slide.)
20 Background-wise the work that has been done up 21 to now by IDCOR for reference plants and by NRC for the 22 six reference plants did not include external events.
23 These were based on internal events only. There are some 24 little overlaps, but la general it did not include 25 external events.
8
a 160 1 At the same time, it does not mean that there is 2 no ongoing work on external events. External events for 3 those plants which are especially exposed to some external 4 events, like plants exposed to high winds or plants which 5 would be exposed to external floods, have been considered 6 in the licensing process.
7 In addition to that, we have a relatively large 8 seismic research program which has been ongoing now for a 9 number of years and that considered seismic events and the 10 effect of seismic events on plants.
11 Similarly, the industry has their own programs 12 also in the various areas and they have various
,, 13 committees, owners groups and review committees that I i
-2 14 overlooks the industry's seismic programs.
I5 In order to accomplish the main goal of the 16 policy statement to provide stability in licensing 17 relative to severe accidents and to be able to tell 18 utilities what will be required in terms of severe 19 accidents and assure them that no more than this will be 20 required from them, one obviously has to look at all 21 severe accidents, and that includes external events also.
22 So we believe that external events should be 23 considered in the implementation of the severe accident 24 policy statement.
25 We looked at also the analytical techniques 8
161 l
1 available for external events and seismic events and we 1 2 found that especially for seismic events they are in 3 reasonably similar shape as it is for internal events.
4 There have been many PRAs done by now for 5 external events and there has been methodology developed 6 for detailed analysis, and both the industry and NRC is 7 developing simplified methodologies for seismic events.
8 Some of these are somewhat behind the internal 9 events like, for example, the application of the 10 simplified methodology for seismic events just will be 11 starting sometime this year, while for internal events it 12 has already been almost completed.
,- 13 So based on this background we have developed a la program and we have a proposal of how to proceed with the 15 external events. We would like to bring this to your 16 attention, especially since the committee has expressed on 17 a number of occasions a rather strong feeling that is external events should be included in their evaluation of 19 the severe accidents.
20 MR. MICHELSON: Is a pipe break outside of 21 containment considered an external event?
22 MR. ROSZTOCZy: No, I don't think so. Internal 23 would mean anything basically that initiates from the 24 plant as opposed to some external family.
25 MR. MICHELSON: A fire though is considered 8
162 y
4 1 external even though they are initiated internally in the 2 plant?
3 MR. ROSZTOCZY: Right. There is fire from 4 internal sources and fire from external sources. For 5 example, the IDCOR methodology as it stands in general did 6 not external events. It does include fires which initiate 7 inside the plant, but it does not include fires external .
8 of the plant.
9 (Slide.)
10 So what is there in this recommended approach.
Il The recommended approach is that we believe external 12 events should be considered. However, we don't think that
,, 13 the schedule for the external events and the extent of the l
- " 14 review for the external events necessarily have to be the 15 same as for internal events.
16 So we are proposing an approach that would 17 proceed on a somewhat slower schedule than the internal 18 events, and only after the first phase is completed would 19 we determine what the extent should be.
20 We believe that this is probably the best to 21 follow at the present time because we don't see any real 22 urgency on the external event. If anything, our judgment 23 is that the risk from external events is smaller than from 24 internal events. And also from a practical viewpoint, we 25 find it difficult to proceed with both of these on a very 8
J t
163 A. ..
\s,/ 1 tight schedule, the type of schedule that I outlined to 2 you here today. So because of that we are recommending to
- i. 3 do this somewhat slower.
4 We are also recommending that the effort should 5 concentrate on plant vulnerabilities. -
6 Another question which is frequently raised, the 7 question ^of what is the contribution of external events to 8 the overall risk, is a difficult question and not easy'to 9 answer. We feel that it is maybe not fruitful at the ,
10 present time to try to concentrate on that and, instead,
- 11 we recommend to concentrate on external event
$ 12 vulnerab'111 ties in the plant rather than trying to answer i
13 that question on the risk contribution.
I
] 14 The first phase of our program, what we are 15 Proposing, then would look at all existing programs and j 16 Programs which'have already been completed and see what 17 have they done in terms of establishing certain standards 18 for external events. This would be done for seismic as 1 -
} 19 well as for other external events.
i l 20 We would also estimate the margin, what these
- 21 Programs provided. For example, if the requirement that 22 we are presently enforcing for high winds, like tornadoes, 23 if those have very large margins in them, then we might
\
24 find that it is not fruitful to try to do some additional 25 work in that area and then we would exclude those from Ill
164 r~~1
(_) I further considerations. I l
2 At the same time, if we find that some areas 3 have looked at that kind of design basis type of event but 4 they have not looked at all of the events that we think 5 are appropriate to consider under the severe accident 6 program, then we would recommend that an additional search 7 for vulnerabilities in that area should be conducted.
a We would also as part of this program establish 9 the limit of how far is it necessary to go. Severe 10 accidents are infinite in the since that we define severe 11 accidents as more severe than design basis accidents and 12 that can go forever.
} ,
13 However, we want to concentrate on that end of 14 the spectrum where we expect that the most return can be
.! 15 obtained for our work. So we want to work on those severe 16 accidents which have a relatively large likelihood.
17 Once the first phase of the program is 18 completed, then we will be in a much better situation to 19 define exactly what needs to be done for individual plants 20 and in which areas one ought to do some additional work, 21 and then that part would be done in the second phase.
22 We don't have an exact schedule for the 23 program. Basically our goal would be to try to accomplish 24 the first phase this year and t'Aen go into the second 25 phase from there on.
i.
>, 165 1 That completes my presentation.
l 2 MR. KERR: Further questions?
3 (No response.)
4 I-turn things over to you, Mr. Chairman. j 5 MR. WARD: Very good. l 6 Let's see, we still have two items ahead of us.
7 Bill, let's see, will you be wanting the 8 committee to do anything at this meeting?
9 MR. KERR: We have been asked to write a 10 letter. I have the draft letter which I will distribute 11 which you can look at over the evening and we can consider
~
12 it tomorrow.
13 MR. WARD: Okay.
- 14 MR. KERR: I am correct, you did ask for a 15 letter?
16 MR. ROSZTOCZY: That is correct. We would like 17 to hear the committee's views on the implementation la program. If the committee feels that there are any areas 19 that we might not have addressed in our program, then we 20 would like to hear about that, of if you feel that in some
- 21 areas maybe we are doing too much and maybe it is not 22 worth to do that much, we would appreciate that also.
23 MR. MICHELSON: What does the letter look like?
24 MR. WARD: He hasn't passed it out yet.
1 MR. MICHELSON:
'25 Oh, I am sorry. Excuse me. I Ill .
l
,. l l
166
?"7 (s,J l thought it was passed out.
2 MR. WARD: Well, we have two more topics. Dr.
3 Carbon has the Advanced Reactor Subcommittee report, and 4 then following that Dr. Moeller has a brief report he 5 would like to give us on some aspects of the TMI-2 6 cleanup.
7 Let's take a break until 5 o' clock and come back 8 and continue.
! 9 (Short-recess.)
10 MR. WARD: The next topic is the advanced 11 reactor designs with a report of our subcommittee by Dr.
4 12 Carbon.
.,,y 13 MR. CARBON: As you know, DOE and its
! 16 interacting with DOE at an early stage in the design 17 effort, and the ACRS has been asked to become involved in 18 a review effort also at an early stage.
19 The meeting today is intended for DOE and its 20 contractors to present the broad outlines of their HTGR 21 design to you, and we have asked them to come back next 22 month and talk about their liquid metal reactor design 23 efforts.
1 24 We are not being asked to write a letter at this 25 meeting.
1 1
167
- 1 We are about 10 minutes ahead of schedule and I 2 would like very much to stay on that. Therefore, my 3 introduction is going to be very brief, and I have been 4 assured by Mr. King of NRR and Mr. Gavigan of DOE that 5 they will be brief also. We will then aim toward a full 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of technical material on the HTGR design.
7 In closing my comments, I would like to suggest 8 that you be alert for several specific safety topics, 9 particularly, one, the temperature at which fission 10 products are released from the fuel, the second, the peak 11 fuel temperature in worst case accidents, the third, the 12 type of ultimate decay heat removal system they have,
,, 13 four, questions about graphite fires, five, the need for
\' 14 containment and, sixth, the need for lack for an emergency 15 preparedness plan.
16 Finally, I would report that many of the 17 important major utility presidents in the country appear is serious about gas cooled reactors, and they state that 19 they are willing to put some of their own money into a one 20 module demonstration plant, and I think that perhaps is 21 meaningful.
22 Let me then call on Mr. King from NRR to speak 23 briefly about their aspect.
24 MR, REMICK: Max, when you say put their own 25 money, I assume you mean their own utility's money?
8
168 7,
_; 1 MR. CARBON: Yes.
2 (Laughter.)
3 MR. SIESS: I believe they have put money into 4 the GCRA for quite a while..
5 MR. CARBON: yes, and I have a letter here which 6 the utility presidents have sent to several Congressmen in 7 which they say that they are putting in utility money.
8 MR. KING: My name is Tom King. I am with the 9 NRR staff.
10 (Slide.)
11 In the old organization I was the Branch Chief 12 of the Advanced Reactors Group. In the new organization I
,-- 13 am a Section Leader in the Safety Program Evaluation h5 14 Branch, and this branch has all the advanced reactor work 15 that is going on in NRR.
16 As Dr. Carbon said, the primary purpose of 17 today's meeting is to acquaint you with the HTGR design 18 that DOE and its contractors are working on and that we 19 have been interacting with them on for approximately the 20 past year to 18 months.
21 The work and its interactions has been under the 22 envelope of the Advanced Reactor Policy Statement which 23 was issued for public comment about a year ago. It still 24 is not out in final form, but we are proceeding to 25 interact under the words and the guidance that are in the 8
169
, 1 version that was issued for public comment.
2 Basically the advanced reactor program consists 3 of looking at two designs, an HTGR design and two liquid 4 metal reactor designs. They are all sponsored by DOE. We 5 have had a number of interactions on key issues and our 6 plan was to review in FY 86 key issues and then in FY 87 7 and into FY 88 review what we call a preliminary safety 8 information document and write an SER on the conceptual 9 design.
10 The preliminary safety information document 11 would also have associated with it a probabilistic risk 12 assessment based on the conceptual design.
-, i3 We had estimated our staff resources as
(,
14 approximately taking five to six staff people per year and 15 about one and a quarter million dollars technical 16 assistance per year, and these resources were all 17 allocated from NRR.
18 Recently about a month ago due to the Gramm-19 Rudman impact, we have reduced our effort. We are now 20 down to about two staff years per year with no technical 21 assistance money. So at the current time we are 22 reassessing what we can do with this reduced level of 23 funding, and we will be having discussions with DOE to 24 take a look at where should we go from here and what 25 should we concentrate on.
8
170
) 1 MR. EBERSOLE: Let me ask a question. Back in 2 the old days such reactors were sponsored not by DOE or '
3 AEC but by individual manufacturers, GE, Westinghouse, et 4 cetera, et cetera, et cetera. I understand this is not 5 the case now, right? You are the sponsors, DOE.
j 6 MR. KING: DOE is the sponsor, that is correct. 1 7 MR. WOODS: With is with NRR.
8 MR. EBERSOLE: Yes, I know. I am talking about
! 9 DOE is. But in DOE there is no particular vendor or 10 designer who is kind of ---
11 MR. KING: Yes, there are particular vendors, 12 and in fact there are some utility industry contributions 13 to this work. I think when Frank Gavigan from DOE follows
- 14 me, he would be the one to address those various 15 contributions.
6 16 MR. EBERSOLE: All right. Thank you.
17 (Slide.)
18 MR. KING: Currently if we have to maintain our 19 reduced level of effort on advanced reactors, we will i 20 probably concentrate on some key issues, and I just put up 21 a few examples of what we are talking about and what we
! 22 mean by when we say key issues.
! 23 The question of containment versus confinement, j 24 the design you will hear about today has no containment 1
25 building.
I i
171 1 The question of using interface criteria only 2 for the balance of plant, the proposals we have heard so 3 far, we would like to, when they send in for a standard 4 plant review, not include the balance of plant because it 5 is not safety related and they feel there is an adequate 6 set of interface criteria and that would be sufficient for 7 the NRC review and approval.
8 The use of a single passive, high reliable, and 9 I left the words out, safety grade decay removal system 10 versus the redundancy and diversity type requirements we 11 have today.
12 And how these plants would treat severe
,_s 13 accidents in emergency planning you will hear today from la the HTGR. They have a lot of inherent passive, highly 15 reliable decay heat removal and reactor shutdown and they 16 would like to have to some relief in the emergency 17 planning area.
18 MR. EBERSOLE: I guess that also has the magic -
19 word " dedicated" that you left out.
20 MR. KING: Yes, dedicated.
21 MR. LEWIS: Does the word " passive" simply mean 22 no electrical power or does it mean more than that?
23 MR. KING: No electrical power and no valves 24 have to open or close, no mechanical operator actions have 25 to take place.
8
_ _ - . _ __.m . . _ . - _ . ._ - _
4 l-172 1 - 1 MR. LEWIS: No moving parts?
2 MR. KING: No moving parts, except for the y 3 fluid.
4 l (Slide.) j i
i 5 This is just a quick summary of what our 6 accomplishments were, major accomplishments to date. We
- r
- 7 have had meetings with HTGR on their top level design i <
8 criteria, their approach to emergency planning and their k 9 overall design approach.
10 On the liquid metal reactor we have gotten into j li several specific issues, their metal fuel, safeguards and l 12 security application of light-water reactor and generic
. 13 safety issues.
i 14 We currently have three major things under j 15 review which we would plan to complete, the HTGR top-level 16 criteria, several reports on the liquid metal reactor 17 shutdown heat removal and a report on the liquid metal 18 reactor safeguards and security plan. i
) 19 That concludes my remarks.
20 MR. REED: your preceding slide was very
! l
] 21 interesting to me. Of course you are applying it to gas
{
22 cooled reactors. But, you know, if you' apply those same j 23 key issues to the light-water reactor PWR that we have
- 24 today, in other words, you go out and decide what the l
l 25 elements of a safety envelope are for the pressurized
!111 k
4
173
_j 1 water reactor, you could set those things aside and 2 regulate the hell out of those and say the rest of the 3 plant doesn't need to be regulated and you begin to come 4 into the same competition as you are probably thinking 5 about here.
6 MR. CARBON: Thank you, Glenn.
7 Move on.
> (Laughter.)
9 MR. REMICK: There is a concern apparently or a 10 need for discusolon on approach for emergency planning on 11 the HTGR. Is this related to the fact that there is no 12 containment? Why would this be different than an LMFBR or
,_ 13 a LWR 7 Why for the HTGR was there need for discussion on
\ )
14 emergency planning?
15 MR. KING: Well, I think this approach, 16 depending on the design, could be applied to any type of 17 plant, but the HTGR approach basically is to try and limit la the off-site doses at the site boundary to less than the 19 protective action guidelines. They feel if they do that, 20 then they would not have to come up with an emergency 21 evacuation plan.
22 The question is what accidents do you have to 23 look at to evaluate the design upon to make sure it does 24 not exceed the protective action guidelines.
25 What that bullet means is they have come in and 4 8
174 1 given us their presentation on this approach that they 2 would like to take. The staff hasn't taken any position 3 on that at this point.
4 MR. REMICK: But that discussion would apply to 5 any type of reactor if they were proposing something like 6 that; is that correct?
7 MR. KING: Correct.
8 MR. MOELLER: On the LMR, what does it mean 9 metal fuel?
10 MR. CARBON: Could we hold that for next month?
Il MR. WARD: That is a long time to wait.
12 (Laughter.)
13 MR. MICHELSON: Liquid metal cooling.
14 MR. MOELLER: Well, does it mean 11guld metal 15 cooling?
16 MR. CARBON: It is metal fuel instead of oxide.
17 MR. MOELLER: Oh , instead of oxide, okay.
18 MR. KING: I am finished.
19 MR. CARBON: Any more questions?
20 (No response.)
21 Thank you, Tom.
22 I will call on Mr. Gavigan from DOE.
23 MR. GAVIGAN: The presentation today is by DOE 24 and a number of DOE contractors, GA, Combustion, Bechtel 25 and General Electric.
8
175 1 I manage the Advanced Reactor Program in DOE.
2 We are doing two LMRs and one HTGR.
3 The difficulties were, and let me just very 4 quickly give this to you.
5 (Slide.)
6 What we are trying to do is develop, as you can 7 see, a low-cost, passively safe generating option that is a competitive with contemporary alternatives. A difficult 9 job, no doubt about it. We have to do a lot of innovating 10 things and you will hear more about them today and next 11 month. Both LMRs and an HTGR are involved in it.
12 (Slide.)
~s 13 Just a very quick look at a preview. Note that i
la we have gone to, and here is the equivalent electrical is rating per reactor, they are somewhat small, and you might 16 call them intermediate sizes, tied up in a modular fashion 17 with turbine generators and putting banks of these on a 18 reactor site to fulfill the needs of a particular utility.
19 This is happens to be an LMR, an LMR and the one 20 you are going to hear about today is this combination 21 here. It is in your handout.
22 (Slide.)
23 I don't have to tell you about the problems of 24 the industry, but I want to quickly let you know that we 25 have conducted a number of survey with utility people to 8
176 n
k_) I try and find what they perceive to be the problems that 2 they are facing, and we in turn wanted to address our 3 reactor designs at solving those problems.
4 None of these are new to you, I am sure, but 5 generally they are talking about uncertainty in cost, 6 uncertainty in safety, risk that the utilities face and 7 difficulties with NRC itself.
8 The characteristics of these designs, what you 9 will hear today and again next month. Generally the tend l 10 to be the small and intermediate size and modular in the 11 sense that they can be added to a site as a function of a 12 load growth demand curve, whatever it might be for a 13 particular utility to make it easier from the viewpoint of 14 financing and to r.ake it easier from the viewpoint of the 15 risk that is involved and the money that is put down for 16 any particular design.
17 We are emphasizing passive safety. We are 18 looking for certified and standardized plants. We are l 19 emphasizing shop fabrication to reduce onsite fabrication j 20 with all the advantages that come with that. They will be i 21 preassembled in effect, barge or rail transportable, cost i
22 competitive as a result of some of these steps we are j 23 taking here, very short construction times, they are 24 looking like 36 to 48 months onsite construction times, 25 and high plant capacity factors.
111 i
i
177 1 Yes, sir.
2 MR. REED: You know, all is not gloriful in shop 3 fabrication, because what you do when you create small 4 modules and shop fabricate is you make things small and 5 crowded. We are all familiar with packaged waste disposal 6 evaporators as an example, which are a horror to maintain, 7 to access and to operate. I am not so sure that shop 8 fabrication is a panacea. Are you?
9 MR. GAVIGAN: I am not saying it is a panacea 10 and I am not saying I am sure. I am saying that so far 11 from what we have been able to see, these things are 12 fabricable in large factories, that we have stable work
,_s 13 forces, we get learning curves that we wouldn't otherwise 14 see and we get QA control that we wouldn't otherwise see.
15 So there are some cost advantages. They are 16 offset by disadvantages of cost, which is the cost of 17 transporting these things to sites and the idea of boxing 18 them in effect in new ways. That adds additional cost.
19 This is doable. It has been done throughout the 20 world for large Ken plants. We are looking at it because l 21 it seems to offer one advantage of, namely, very short 22 onsite construction times.
23 MR. REED: Are you familiar with such as a 24 package waste evaporator manufactured by a manufacturer 25 many years go and these kind of things and how they 8
178 r- -$
_j 1 perform in the field and how difficult they are to 2 maintain?
3 MR. GAVIGAN: No.
4 MR. REED: They were modules, too.
5 MR. GAVIGAN: Well, I don't think you can be 6 against the idea of modules. You talk to the Bechtel man 7
later on and he wi]I tell you about how they fabricate 8 wholesale Kem processing plants and mail them in effect to ,
9 New Zealand and they perform onsite.
10 I think it is not the module itself, but perhaps l 11 it may be how it was designed and how it was put together l
12 and the quality control it waa inunluad in.
ry-13 MR. REED: I don't think we are comn.unicating 14 very well. As a person who has worked in the field'and 15 seen field erection and field space and these kinds of 16 things in nuclear plants and recognizing the problems of 17 radioactivity and shielding and all these things, I would 18 be concerned about modular construction.
19 MR. CARBON: You had better move on.
20 MR. GAVIGAN: Yes.
21 MR. LEWIS: I wonder if I could ask --- and I 22 hate to do it, Max, but if I could ask one other 23 question. I am intrigued by the idea of mailing a 24 reactor, because that is a new challenge to the Post 25 Office to mail a reactor.
8
l 179 1 (Laughter.)
2 Passively safe is a kind of bandwagon these 3 days, and I wonder whether you are confident that 4 passively safe is better than perhaps with a dedicated 5 battery or something like that, that the passively really 6 contributes reliability to the safety system?
7 MR. GAVIGAN: So far from what we have been able 8 to see, it certainly does. There are arguments on what is 9 passively safe andfwhat is inherently safe and so on, and 10 we are addressing that. As a matter of fact, I have a 11 special viewgraph but no time to spend on that, but that t? in wnvth talking about.
-- 13 It gives generally simplicity. It eliminates a
14 lot of hang-on systems and it allows you to understand how 15 a system behaves. It is much more simple and you will 16 hear that today.
17 MR. LEWIS: Okay, fine. I will wait until then, is but the passive schemes I have seen have seemed a little 19 more complicated than the known passive ones.
20 MR. GAVIGAN: This is certainly not true with 21 what you are going to be seeing today.
22 MR. LEWIS: Okay, I will wait.
23 (Slide.)
24 MR. GAVIGAN: These of course you are familiar 4,
25 with, the advanced reactor policy statement requirements.
8
180 7y
( _j 1 You will notice that the requirerents that one must meet 2 to become an advanced reactor are the those very 3 characteristics and features that we have in our existing 4 plants which I showed you on the viewgraph before.
5 (Slide.)
6 And the last viewgraph, where are we. We are 7 part way through the conceptual design. The LMRs,are 8 slightly ahead of the HTGRs in time because we started 9 earlier. We have had a positive experience with NRC.
10 They are open minded and it is very useful and interesting 11 for them and us.
12 Tha preliminary cost ectimatec 00=c in that all r- 13 three of these plants are competitive with LWRs and I
( >I 14 capital costs, and all of them beat coal on operating 15 costs.
16 We have enhanced passive core shutdowns, heat 17 removal processes, and these are all natural circulation 18 generally, non-moving parts, depending on natural laws 19 primarily. We have R&D needs identified from both the 20 HTGR and the LMR concepts, and they help folks to 21 supporting R&D programs. So all those resources are 22 mobilized to answer these questions that you are raising.
23 And, lastly, we have the utility interaction.
24 You know about the GCRA activity, and the two LMR 25 designers are begir.ning to build that kind of interaction 8
I 181
,/ 1 and support for their own designs.
2 With that, I will let Andy Millunzi come on, 3 unless there are any more questions.
4 MR. REMICK: I believe that DOE and DOD are 5 working on small modular facilities, too. Is there any 6 relationship between that program and this program?
7 MR. GAVIGAN: Not direct. The sizes they are 8 working on are 10 megawatts electrical, for example. We 9 are talking here about 150 to 350 electrical.
10 (Slide.)
11 MR. MILLLUNZI: Here is the agenda that we will
!? be fc11cMing for the rect of the time. I will try to keep
, 13 my remarks very brief and hope to be off in less than 15
(-' )
u minutes so that we can concentrate on providing you an 15 overview on the design of our high-temperature gas reactor 16 and then what the safety characteristics of that reactor 17 are. Both of those discussions will take 25 minutes.
18 MR. CARBON: If you can cut yours to about five, 19 we will appreciate it.
20 MR. MILLUNZI: I will try to very much.
21 I think it is important to follow up a little 22 bit on the letter that Dr. Carbon read from the 23 utilities. Here are the participants in the gas reactor 24 program. Our support from the utilities is very broad, as 25 he said.
8
182 P -q 1 Today we have with us Mr. John Rucknagle from 2 Public Service Electric and Gas. There are going to be 3 about two or three, and I don't know the exact number, of 4 other utility personnel who were going to be here, but 5 this weather socked them in and not that it was Friday. .
6 (As Dr. Carbon has requested, I will go through 7 my presentation very fast.
8 Normally I would think that my presentation no 9 would give. However, we have found it absolutely 10 necessary that we at least give you a brief overview so 11 that you know how we are entering into this second I? generntion of reactors.
7.-- 13 The comments that Mr. Reed has raised and some i
'" 14 of the other questions that come out I think really stem 15 from all of our very deep understanding of the present 16 light-water reactors, understanding that those are safe 17 reactors and that they are viable machines. Therefore, l
18 the second generation has a very, very tough job to go in 19 trying to develop something better.
20 The reason I need to talk to you is that we have 21 a program which we are very proud of in that it is a 22 cohesive, comprehhnsive and a very rigorous program. We 23 think that it is necessary for you to understand that we 24 are coming from the top down and that the program is 25 driven by requirements. So you need to understand that 1
183 1 because then that makes the design understandable of why 2 the technology programs are in place to meet those 3 designs.
4 So we had a very fruitful meeting with the 5 subcommittee which took six and a half hours, and today we 6 are trying to compress that down into an hour. So I will s 7 move on as far as I can here so you can get to the other 8 parts.
9 (Slide.)
10 I will skip over several of the viewgraphs. I 11 will put up again the objective and just tell you that the 12 HTCn reactor is tle unique reactor in that it is not only 13 applicable to electricity generation, but it is also I ,I 14 applicable to supplying other energy sources.
15 In fact, the capability of the HTGR to meet 16 these other energy sources will provide an energy option 17 twice as large as if we meet all the electricity. So it la is a very worthwhile national goal.
19 (Slide.)
20 You have heard about our interactions with the 21 NRC. Here is our schedule of activities which is in the l 22 handout. In summary we will have 31 interactions with the 21 NRC staff between now and the end of fiscal '87. Of that 24 there will be 21 meetings and that does not include our 25 anticipation of having to respond to questions that the 8
I
184
_j 1 staff will have when we submit our preliminary safety 2 information document.
3 The details of our licensing plan which has been 4 submitted to the Conmission and has been approved by them 5 is in your handout.
6 (Slide.)
7 The next viewgraph describes what our design and 8 licensing approach is. As I said, we come from the top 9 down. It turns out that from the top down means that we 10 are designing this plant to meet both user requirements 11 and the top level regulatory requirements.
12 Taking those requirements, we are using a very 13 disciplined rigorous framework by which we can understand j
l }
'" 14 what functions have to be performed to meet those I i
15 requirements, and from those functions and those 16 requirements we are developing the design.
17 Now we are in a second generation of a reactor.
18 These reactors have features which are different than the 19 present system. So it is necessary for us to develop a 20 methodology which will translate this advanced approach 21 into a format that the Commission is familiar with l 22 dealing.
1 23 (Slide.)
24 Today I will very briefly run through on this 25 part of the blue diagram, very briefly on the yellow and 0
185
- 8
_) i very briefly on the licensing basin.
2 Tom Neylan will give you a 25-minute description 3 of this design, and Fred Silado will give you a 25-minute 4 description of the safety characteristics of this plant.
5 Not to put a challenge, but to re-emphasize, I 6 indulge in asking you to give us an open mind and listen 7 and see how we have been trying to be innovative and truly 8 coming into a second generation of reactors.
9 One of the things that I want to emphasize as we 10 go through again is remember that the design that you hear 11 is going to be based on these requirements and this I
i2 opproach. We do not have time to talk to you today, but I
<s 13 hope today that we spark your interest enough that we
' ~
14 would have the opportunity to come back and to talk to 15 those in even more detail.
16 (Slide.)
17 Here is a brief description, not the full list 18 of the utility user requirements that we are designing 19 to. These requirements have been approved by the Chief 20 Executive Officers of one-third of the electrical 21 generating capacity utilities in the country. It is 22 important to look at this equivalent availability.
23 I know, as Mr. Reed, being from Wisconsin knows 24 how interested I am about forced outages and getting 25 reliability. This 80 percent is divided up into two 8
186
,J l separate requirements.
2 One is that the plant outage only be 10 percent, 3 the forced outage only be 10 percent and they can't 4 combine those in order to get to the 80. We will not 5 accept 15 enforced outages and five planned in order to 6 meet the 80. The reason for that is if we can get this 7 design to meet the forced outage, it means we have a 8 highly rollable plant that is easy to operate and, very 9 importantly, very easy to maintain.
10 Although these plants are modular, we have 11 started out from day one,.'as you will see here, thinking 12 of the guys who operate and maintain, and our biggest
,, 13 question is how is the guy out in the field going to take
' -) 14 care of this plant.
15 All of this, let me say, we are benefiting from 16 the experience that has come out of the light-water 17 reactor which has been operating very well as a reactor.
18 (Slide.)
19 Now the next thing we had to do was to develop 20 the top-level licensing regulatory criteria. And when you 21 get there you say what are the bases.
22 These are the bases that are here. I think, as 23 you see in aummary here, that it is necessary that they be 24 direct statements, that these criteria be independent of 25 the design because to the public they don't care where 8
r 1
1 187 I that is coming from. So the criteria must be independent 2 of the design, and most importantly, they must be 3 quantifiable so that they can be measurable and not end up 4 in subjective arguments which have been plaguing the LWR 5 industry.
6 (Slide.)
7 Now our proposed sources for these top-level 8 regulatory criteria come from these documents.
9 Now we come to the basis, if you recall on the 10 viewgraph, the blue part is the integrated approach.
1: What is this integrated approach? What we have 12 done is we have gone back to square one and we asked 13 ourselves the question why the hell are we in business and la why are all of us here involved in the nuclear power 15 business, and that is because our objective, our top 16 objective for all of us is to produce safe, economical and 17 reliable power using the nuclear fission.
18 What we have done is we have developed a very 19 rigorous framework to do this.
20 (Slide.)
21 And coming from the top down, what we have here 22 is the safe economic reliable power. From that we have 73 two sets of requirements, both the regulatory and the 24 utility user requirements.
25 Now it is important as we go into this second 8
188
(_s 1 generation of plants that we re-establish the correct 2 roles and responsibilities between the applicant and the 3 regulatory bodies.
4 Because the utility user must meet both sets of 5 those requirements, he must be free to allocate his 6 resources as he sees fit to do that. That means that the 7 roles are to be that the requirements from the regulatory 8 side get set by the NRC acting as the agent of society, 9 and then that the utility, the applicant determine how he 10 is best going to meet those requirements and that then his 11 proposed way be evaluated against these criteria of the 12 kind that I mentioned to see if they are good enough.
l 13 What we want to resist is getting into the 14 business where people tell us how to design the reactor, 15 and we would welcome suggestions, but we hope nobody gets 16 personally offended if in our evaluations we find another 17 way to go better. What we would like to have you evaluate 18 is in the way we did it, is it good enough.
19 (Slide.)
20 Now in this approach we take these requirements i 21 and we allocate them down. This frame, which is 22 represented here, which I don't have time to go into 23 details, is what has to be done. In response to what has 24 to be done, there are four institutions, the design, the 25 construction, operation and maintenance.
8 i
189
_j 1 And what we have done from the very beginning 2 has to integrate these things all into one so that from 3 the beginning the guy who is going to end up operating and 4 maintaining has his input to the designer and vice versa, 5 the designer and the constructor is able to tell the 6 operator and the maintenance why he did what he said he is 7 going to do.
8 MR. CARBON: We are running out of time.
9 MR. MILLUNZI: Okay. I will skip the next 10 viewgraphs which are a development of this process.
11 (Slide.)
12 In summary, one of the things I would like to
,_ i3 say is that our safety and licensing approach is
( )
la consistent with the advance reactor policy statement and 15 these bullets are just in summary.
16 With that, I will turn it over to Tony Neylan 17 who will describe the design.
18 MR. NEYLAN: Good evening, gentlemen.
19 I am Tony Neylan. I am going to give you a 20 brief overview in the next 20 to 25 minutes or so of the 21 HTGR characteristics, and I will be followed by Fred 22 Silady who will describe the safety characteristics of the 23 Plant that I describe to you.
24 (Slide.)
25 I would, first of all, like to start with a 8
e
190 1
J 1 viewgraph that identifies the key characteristics of the 2 HTGR and the basic characteristics of all HTGRs.
3 (Slide.)
4 First, we use helium gas as the coolant. Helium 5 is of course inert, it is a single phase and does not 6 change phase with temperature and because it is a gas, we 7 do not lose coolant. We simply depressurize.
8 We have a graphite moderator, and because of its 9 very large thermal capacity, we have long response times 10 to abnormal events.
11 The core is supported by a graphite core support 12 structure which maintains its structural stability. In
, 13 fact, it increases in strength with temperature up to the i i td 14 temperatures that we will be talking about.
15 A key in the design is that we use a ceramic 16 coated fuel particle embedded in a fuel raatrix in a 17 graphite block as at Fort St. Vrain, and that has resulted 16 in very low releases to the operator, low releases from i l
19 the fuel particle itself, and I will be talking to you in l l
20 a little more detail about that shortly, and Fred will l 21 show you the specific releases on the temperature 22 conditions.
23 Now we have taken these basic features and 24 developed some spes:lal modular features, the basic 25 characteristics which are that we have selected a 8
191 j 1 configuration, which because of its size by limiting the 2 power density of the core and by keeping the configuration 3 of the core such that the surface to volume ratio is such 4 that the temperatures are limited and we have provided a 5 heat removal mechanism that does not rely on any active 6 components, we are able to assure public safety, and you 7 will again see this in some detail in the subsequent 8 Presentation.
9 We anticipate that we will develop a 10 standardized design, that we will get at pre-license and 11 that certain major elements of it will be factory 12 assembled and shipped to the field.
7- s 13 MR. EBERSOLE: May I ask a question before you i
)
14 put that down?
15 MR. NEYLAN: Certainly.
16 MR. EBERSOLE: You said those were the key 17 characteristics, but you didn't mention the thing that is makes the Fort St. Vrain reactor an operational disaster, 19 which is that business of using water buffered seals that 20 Persistently pump water in onto this hot environment. I 21 should have thought that would be a key difference.
22 MR. NEYLAN: It is indeed a key difference in 23 this design. However, here I was addressing the physical 24 characteristic of HTGRs and their benefits toward public 25 safety.
8
192 r,
_j 1 I was addressing here those configuration 2 selections that we have made which are unique to this 3 design which assure public safety. And, indeed, although 4 Fort St. Vrain has been plagued by water ingress and has 5 suffered a terrible availability record because of it, and 6 it has simply large been that that it has been shut down, 7 we intend to cure that by changing from the steam driven 8 circulator to an electric motor driven and eliminating the 9 source of water to the bearings and the present design 10 carries magnetic bearings.
11 MR. EBERSOLE: Thank you.
12 (Slide.)
_- 13 MR. NEYLAN: This is an artist's representation I l 14 of the plant. The plant size is about 1500 feet by 900 l
15 feet occupying about 26 acres. It is sited on a site 16 having an exclusion area boundary of 425 meters and 17 encompassing about 140 acres. For the reference 18 configuration we have four modules and two turbines.
19 There are two points I would like to stress on 20 this particular viewgraph. One, we have located the 21 modules, and this is the area I will focus in, below 22 grade, and then we have a conventional structure above 23 grade which in fact is a service and provides cranes and 24 so forth with access to the modules, but it is not a 25 containment. It is in fact a very low profile plant i l
193 x_ i because of this configuration.
2 To focus in on the module itself, we have two 3 separate vessels, steel reactor vessels using light water 4 reactor technology. One vessel contains the reactor core.
5 It is approximately 22 feet in diameter and about 70 feet 6 long. It weighs about 700 tons.
7 The steam generator vessel is about 14 feet in 8 diameter, it is about 85 feet long and it weighs about 300 9 tons. They are connected by a concentric cross-duct, and to you will see in a moment the helium flow path on a later 11 viewgraph connecting these two reactors.
12 There are 18 control rod drive penetrations in
,_ 13 the top head extending about 16 feet from the top head of I )
14 the vessel to grade level.
15 The main circulator is located in the cold leg 16 above the steam generator, and we have a shutdown heat 17 exchanger for maintenance purposes and its own independent is shutdown circulator located at the bottom of the reactor 19 vessel.
20 MR. REMICK: One of your earlier slides said 500 21 megawatt electrical and this says 350. The 500 must 22 before the four modules, is that it?
23 MR. NEYLAN: That is correct. It is 350 24 megawatts thermal with each one generating 140 megawatts 25 electric, and it gives you a net output for four of 560.
8
194
- r~ "'1
(,) 1 (Slide.)
2 This is an elevation. This is grade level.
3 This is about 150 feet deep. The structure is i cylindrical, about 63 feet in diameter.
It shows the 4
5 relationship of the normal steam generator main circulator 6 heat transfer system to convert the heat energy into 7 steam. It shows the location of the shutdown circulator, 8 and also very importantly it shows the location of a 9 reactor cavity cooling system in and against the walls of 10 this reactor structure adjacent to the reactor vessel.
11 In th'e event of loss of this main cooling system 12 or the maintenance system, one would get convective
, 13 cooling under a pressurized condition in this cavity
-s' 14 losing heat to the cavity cooling system in either a 15 pressurized or a depressurized condition. And again you 16 will see some more details on those specific items.
1 17 But I would like to come back, first of all, to la a flow diagram from the main heat transfer system.
19 (Slide.)
20 Let me describe the helium flow path. The 21 helium is pushed out, blown out by the main circulator.
22 It flows around the outside of the coaxial duct, the 23 outside of a core barrel and down through the core. The 24 temperature here is approximately 500 degrees F. It picks 25 up heat in the core, exiting at about 1268 degrees F w r--,y- y y .- , - -y--w -
m.,-----,y--,.-mr,, -- , - , -, ,--,--,---#- ,y. ,. + + + _-..py--
w.m+e p-e--.y --ee.g, gw9w9
195
- I average temperature, which incidentally is lower than the 2 1400 degree F average on Fort St. Vrain. It passes 3 through the steam generator and then completes the loop 4 back to the main circulator.
5 The feedwater of about 400 degrees F enters 6 through a bottom penetration into a helically wound steam 7 generator similar to Fort St. Vrain construction, except a that it is simplified and there is no reheat because of I 9 elimination of the need to bring back the steam for the 10 steam driven circulator that I mentioned before. It is 11 uphill climbing and emerges at 1000 degree F, 2500 psi 12 through a conventional and available turbine generator, c 13 through the main condenser and through a normal feedwater I 1
-' 14 chain to make the loop.
15 Using that heat transfer system for the four 16 modules, and this was raised earlier, we have four at 350 17 and a gross thermal power of 1400 megawatts. The helium is pressure is about 925 psia. We deliver the steam l
19 conditions that I mentioned, and the net electrical output j 20 is 558, giving as an efficiency of just 40 percent.
21 (Slide.)
22 Now the shutdown cooling system which, as I said 23 earlier, is provided for maintenance in the event that 24 there is a shutdown in the circulator or in the feedwater 25 chain, one then would use this shutdown cooling system to 8
l l
l l
l
196
,q gj 1 remove the heat from the reactor to cool it down to enable 2 ready access for maintenance.
3 It is supplied with water at 130 degrees F, and 4 this charges at about 450 degrees F through a conventional 5 heat exchanger. It has its own independent circulator.
6 The flow path for the helium and it will work in either a 7 pressurized or a depressurized condition, is in the same 8 manner on the outside of the core down through the core, 9 but this time through this heat exchanger and then back in 10 the same loop.
11 (Slide.)
12 This is a schematic of the reactor cavity 7- 13 cooling system which, as I pointed out, is on the
- i b- elevation.
14 It is located adjacent to the walls of the 15 silo opposite the reactor vessel.
16 It is ducted cold air in through in the green 17 shown on this viewgraph returning through convected 18 heating through the yellow and to an exhaust duct. There 19 are four redundant exhausts and intakes.
20 It operates under normal conditions to keep 21 ambient conditions within the reactor cavity and it 22 maintains the concrete temperatures around about 70 23 degrees F. It is the system which removes heat from the 24 reactor vessel by radiation and convective heat flow 25 through the air under already pressurized condition.
197 s , 1 There are no valves and no moving parts in this system.
2 The typical heat load, 700 kilowatts when in 3 normal operation, and about 1700 under the pressurized 4 cooldown condition.
5 We now turn to the reactor cross-section which 6 again is a key element in the configuration which enables 7 this passive safety that we are seeking in the performance 8 of this reactor.
9 The action core is an annular configuration to shown by the shaded hexagonals in this diagram. It is 11 about 3.5 meters in diameter OD and 1.6 meters ID. It is 12 approximately eight meters in height. The power density
,, _ . 13 is 5.9 watts per cc.
1 )
14 Outside of this annular region is a reflector of 15 about 1.2 meters, two rows of removable reflectors and a 16 Permanent reflector all supported form a lateral restraint 17 structure from the steel vessel.
18 The central region, there are also hexagonal 19 blocks. These are unfueled.
20 We have two independent reactivity control 21 systems and a control rod system providing 24 outer 22 control rods and 6 inner control rods of the type at Fort 23 St. Vrain. They have a combined reactivity worth of 24 approximately 21 percent delta rho.
25 There are 12 reserve shutdown channels, again 8
198
. is_; I similar to Fort St. Vrain, using small boronated pellets 2 which are released into these channels. It has a 3 reactivity worth of about 15 percent delta rho.
4 This compares to the requirement under the 5 beginning of cycle, which is the worst case of about 13 6 percent required, and if we had the most adverse water 7 ingress, of approximately 15, which combined would give 8 you a 28 percent requirement versus, if you added both of 9 these systems together, 41 percent capability, giving as a 10 margin about 11.
11 (Slide.)
12 I would like to focus on the fuel.
<~s, 13 MR. MOELLER: Did you tell us the enrichment?
k) 14 MR. NEYLAN: I didn't. I was about to, 9
i 15 MR. MOELLER: Okay.
. 16 MR. NEYLAN: I would like to focus on the fuel 17 which has been successfully developed and demonstrated
23 This has developed into the so-called Triso 24 particle on Fort St. Vrain which has performed so well out 25 there in terms of limiting releases. It is in fact a 8
o 199 s
i buffer zone, an inner pyrolytic carbon, a silicon carbide 2 zone and a pyrolytic carbon.
3 Now I would like to pass around, and it may or 4 may not get around in the time, but some particles of the 5 size that we are talking about. We are using this same 6 basic technology and here we are developing a low enriched 7 uranium fuel. It is an an oxi-carbide fissile particle 8 and it is 19.9 percent enriched versus the high enriched 9 that was used on Fort St. Vrain.
10 But the key to the fission products is 11 essentially in the silicon carbide coatings. The 12 pyrolytic carbide coatings help, but they also primarily 13 protect the silicon carbide.
t
!~' 14 The fertile particle is the same technology.
15 The total size of those particles that are going around 16 has an outside diameter of about 800 microm.
17 (Slide.)
la This is an electron microscope picture of that 19 Particle. This shows the kernel, the buffer zone to 20 contain the fission products, the PYC inner layer, silicon 21 carbide, and PYC outer layer.
22 (Slide.)
?3 Now these particles are combined in a fuel rod, 24 which is about a half an inch long, half an inch in 25 diameter and two inches long.
Ill l l
l i
200 7- q
_; 1 Again, let me pass around a characteristic fuel i l
2 rod.
3 These fuel sticks are loaded into holes in a 4 fuel element and sealed. Adjacent gas passages are 5 drilled through this element and the heat is transferred 6 at fission to the gas and heated up, as I said earlier, 7 from 500 degrees to approximately 1270 degrees everage 8 outlet temperature.
9 These blocks are the same kinds of blocks that 10 are used on Fort St. Vrain. They are approximately 14 11 inches across the hexagonal flat and about 32 inches long, 12 and I have a small model here, the genuine sized ones are rm 13 too heavy to carry, that you can pass around. These are i !
'"' 14 stacked one on top of another in the annular contiguration 15 that I showed you before.
16 (Slide.)
17 We have been one other thing with the fuel over 18 the years, and that is improving its quality. And we in 19 association with Hobeg, who is the German company, have 20 been developing improved quality fuel. Quality is 21 measured in terms of a defect fraction of the fresh fuel, 22 how many of those particles of these silicon carbide 23 layers have defects and how much contamination is there in 24 the process when one finally makes a fuel element or a 25 fuel stick.
8
201 (j 1 And you can see here that the defect fraction of 2 Fort St. Vrain we have improved, and in fact we specify 3 for this modular HTGR an improved quality of 6 times 10 to 4 the minus 5 total defect fraction. You can see that this 5 is the same that Hobeg, the German company, has 6 manufactured commercially to the candidate German designs 7 of HRB and Interatom. They have made particles and it is 8 the same basic particle. Their fuel element form is a 9 pebble as opposed to that prismatic block that you see 10 there.
11 MR. SHEWMON: What do you call a defect and what 12 is the burnup at which you determine it?
n MR. NEYLAN: A defect is where we have initially l'\
sI la failed a silicon carbide layer so it is broken.
15 MR. SHEWMON: To no burnup.
16 MR. NEYLAN: No burnup. This is fresh fuel. So 17 you have a very small fraction of that, and you may also la have some contamination in the procese where you get some 19 UCO on the outside of the particle, but this is fresh 20 fuel. The burnup, to my recollection, is at about 9500 21 megawatt ---
22 MR. SHEWMON: Presumably the leakage that you 23 get at the end of that 9500 should reflect how many 24 defects you have in the fuel.
25 MR. NEYLAN: Yes. We have a specification for 8
202 7__
. ; I the fresh fuel and we have a specification for its 2 preferments at the end of life when it has undergone its 3 full burnup, and we have set particles. We have in fact 4 test capsules as well as the Fort St. Vrain operating 5 experience that shows the releases from that fuel, and we 6 are able to test that fuel.
7 MR. SHEWMON: This increases an order of 8 magnitude from initially or a factor of 10 percent?
9 MR. NEYLAN: It is dependent on temperature and 10 you will see the releases. It is approximately the same 11 ler of magnitude as this for the extreme accident 12 conditions that you will see here. So about half of the r ,, 13 fission gas comes from this initial defect fraction and I l
" 14 approximately half from failure due to temperature.
15 MR. MICHELSON: Is this intended to be a throw-16 away cycle?
17 MR. NEYLAN: It is a LEU and it is a throw-away 18 cycle at this point and all the economics are based on 19 that.
20 At this point then I would like to put on a 21 summary viewgraph which is an introduction to Dr. Silady.
22 (Slide.)
23 Having selected this configuration, we wish to 24 show that this reactor will assure public safety and meet 25 the regulatory requirements without operator action and 8
203 j
(*3 .
(_) ,1 without the operation of any powered system and without !
l 2 the need of public sheltering or eyacuation. It will meet 1
'3 the PAG limits and hence the 425 meter EAB.
4 MR. EBERSOLE: Isn't it.true though, however, 5 that the core under those conditions might be ruined? l 6 MR. NEYLAN: No, it is not..
7 MR. EBERSOLE: Oh, isn't it?
8 MR. NEYLAN: No, it is not. In fact, we have 9 investment requirements that we can take this temperature 10 and restart, and you will see some of those temperatures 11 in the.next presentation.
12 Are there any other questions?
13 MR. REMICK: The passive heat removal is 14 radiation from vessel ---
15 MR. CARBON: We must move on.
16 MR. REMICK: Okay, fine.
17 MR. NEYLAN: you will see that in the next 18 presentation.
19 .
20 21 22 23 24 25 8
204 r-,
_ , . 1 EVENING SESSION 2 (6:00 p.m.)
3 MR. SILADY: Good evening.
4 My name is Fred Silady. I am going to be 5 discussing the modular HGTR safety characteristics.
6 (Slide.)
7 The modular HTGR safety design objectives 8 include the following.
9 We intend to provide a fuel quality and specify 10 normal operating conditions that limit the radionuclide 11 inventories that outside of the fuel to such an extent 12 that we can meet 10 CFR 100 doses even if all of those
-- 13 things, all that inventory was released.
("" }
14 Secondly, our objective is to provide a reliable 15 passive design selection to retain the radionuclides 16 within the fuel to meet the 10 CFR 100 doses at the plant 17 boundary.
18 Notice in this second point that the emphasis is 19 on retaining the radionuclides within the fuel with 20 passive design selections which I will be discussing in a 21 couple of minutes.
22 The third point is to provide additional design 23 selections again largely passive to retain the 24 radionuclides within the plant to meet the protective 25 action guidelines at the plant boundary.
8
1 205
'_) 1 MR. MOELLER: Excuse me. I don't follow that.
2 There is a lot of difference between meeting 10 CFR 100 3 and meeting PAGs. So you are telling us that in order to 4 meet the PAGs you need more than you are giving us in the 5 basic plant design?
6 MR. SILADY: The basic plant design has as an 7 objective to meet the protective action guidelines over a 8 very wide spectrum of events which encompass those for 9 which we want to show that 10 CFR 100 doses are met. So 10 this is the limiting one.
11 Our s.3hemis will be on showing that the fuel 12 alone can take care of 10 CFR 100. Some other retention
,, 13 mechanism, such as deposition or plateout or a holdup in
( )
- ' 14 the vessel will help us meet the protective action 15 guidelines.
16 MR. MOELLER: Thank you. That helps.
17 (Slide.)
is This abbreviated functional analysis tree 19 provides a summary of all of the functions that we need to 20 accomplish to meet all of our user and regulatory 21 requirements, things like Appendix I, things like 10 CFR 22 100, but also the user / owner requirements that we not have 23 any offsite evacuation.
24 The shaded boxes in the diagram are those that 25 we are going to rely on in order to meet 10 CFR 100. So 8
_ - _ _ _ _ _ _ ~ _ - -
_l i
206 1
this brings out the point that you just raised.
2 Going through it quito qu2ckly, in order to 3 maintain control of radionuclide release, both of these 4
functions are required for all of our requirements.
5 To control radiation, we are going to emphasize 6 the control of it from the core as opposed to to 7 controlling the radiation from the processes and from s
8 storage where the inventories are quite small. And in .
9
, controlling the radiation from the core, we are going to 10 do it by keeping the radionuclides within the fuel 11 particles. ~
12 We have in the design additional retention ,
13 barriers in order to accomplish the other requirements 14 such as Appendix I for operational occurrences and in is order to meet those protedtive action guidelines.
16 To meet 10 CFR 100, however, it is our objective 17 that we need only retain the radionuclides in the fuel 18 particles.
19 In the subsequent viewgraphs I am going to be 20 emphasizing the retention within the fuel in order to meet 21 10 CFR 100, but carried along in the presentation in your 22 package you will see the corresponding retention 23 mechanisms in order to meet the protective action 24 guidelines.
25 (Slide.}
Ill.
.. . - ~ . . -
207
() 1 I wish to set up a success' criteria by which the 2 other viewgraphs are keyed, and this is converting the 3 construction permit thyroid dose of 150 rem per 10 CFR 100 4 into a release fraction that would be permitted from the 5 plant.
6 This is done by figuring out.the effectivity, 7 the total inventory in terms of iodine 131, which is the 8 . dominant nuclide and the effect of the breathing rates and 9 Chi over Q's from Reg. Guide 1.4 into a fraction of this 10 inventory that could be released that would equal 150 rem, 11 and the other column is a factor of 30 less, the
- 12 difference being 150 rem versus 5 rem.
p 13 (Slide.)
14 MR. MOELLER: Well now in the 5 rem you are is using then the lower limits because it is 5 to 25.
16 MR. SILADY: Exactly correct.
17 MR. MOELLER: You are being very conservative.
18 MR. SILADY: That is right. Well, in both 19 cases. We took the 150 rem at the construction permit 20 stage ---
21 'MR. MOELLER: It is really 300.
22 MR. SILADY: Yes. And we have examined a broad 1
23 spectrum of events that to the subcommittee we went into 24 'how we select those events and the risk assessment.
25 (Slide.)
8 P
, .,-es -,y ,v -
-.m-t vwv- w ,.v- ---w-- , ,*-*-'vtr"vT-*'*M- " ~7" T tW ' - - - - " * - '* '"t***f' ' * * ' ' ' ' * ' *m-F" ~ ' - * * * '
208 (R_,1 -1 We have found that in-this design we can retain 2 radionuclides in the fuel particles by performing these 3 three subfunctions, removing the core heat, controlling 4 the chemical attack and controlling heat generation.
5 So in the remainder of the time I have this 6 evening, I am going to take those in the order I just showed you and look at what are the possible ways that 8 those functions might be accomplished.
9 (Slide.)
10 Before going into retaining the radionuclides 11 within the core, within the fuel itself, this it.s a slide 12 that shows relative to that 3 times 10 to the minus 5 that
, 13 we have limited our fuel quality to such an extent through
"~ 14 the evolution of the information that we have gotten from i
15 Peach Bottom and Fort St. Vrain and the capsule work so 16 that we feel confident that our plateout activity 17 throughout the circuit would only be 80 curies out of that 18 10 to the 7th.
19 The fuel will be that good that during normal 20 operation after 40 years only 80 curies will. plate out 21 around the circuit, and only 80 curies will be available 22 to be released then from the vessel.
23 Therefore, we meet the 10 CFR 100 because that
, 24 is less than the 3 times 10 to the minus 5 even if all of 25 that plateout activity at the end of life were released.
8
~
209
- ) 1 .Also shown on that slide were additional.
2 retention mechanisms that go towards meeting the 3 corrective action guidelines.
4 (Slide.)
5 Now the design selection is to remove the core 6 heat. There are several in this design, and you have 7 heard about the active systems for normal operation, the a heat transport system, and the shutdown cooling system for 9 maintenance.
10 The one I would like to go into more detail on 11 now is the reactor cavity cooling system, and it can 12 operate either under pressurized or depressurized primary 13 coolant conditions.
-O 14 (Slide.)
15 It is the reactor cavity cooling system that we 16 have chosen to rely on in order to meet 10 CFR 100. The 17 reactor cavity cooling system, however, operates in tandem 18 with the following design sections as well:
19 The small thermal rating which limits the amount i 20 of the afterheat, the decay heat.
21 The core geometry which you heard was annular. ;
22 The low-power density comes in in terms of the 23 slow heatup of the massive graphite core.
24 And the fission product retention capability of
]
1 25 the fuel at high temperatures. 1
. 210 1 (Slide.)
i 2 Going through the decay heat modes in just a 3 little more detail than has been done previously, this is 1
4 a blowup of a cross-section where the. main heat transport
- 5 system and the shutdown cooling system have been lost.
6 .The system is pressurized, however, and the heat can be
- 7 circulated, .since it is at 925 psi roughly, up through the
- 8 core and back around just by natural convection.
> 9 At the same time heat is being conducted through 10 the graphite blocks from the middle out. This is the 11' annular core shown here, and this is the non-fueled 12 region, and radiated to the vessel as well as being 13 convected to the vessel.
14 Radiation then occurs from the steel vessel over 15 to the air cooling panels, and there are four of these 16 passive reactor cavity cooling systems with air coming in 17 from the intake going through and out the exhaust.
18 Note that the air is separated from the vessel 19 and it doesn't flow directly over the vessel. There is a 20 steel membrane right there.
21 MR. EBERSOLE: What is the normal temperature in 22 the vessel of metal?
23 MR. SILADY: The vessel during normal operation 24 is approximately 400 degrees.
I 25 MR. EBERSOLE: It is a steel vessel?
111 J
k 211 A
(_) 1 MR. SILADY: Yes, it is.
2 MR. EBERSOLE: What happens when it blows up?
3 MR. SILADY: Excuse me?
4 MR. EBERSOLE: When you blow up the vessel does 5 it carry away the passive structure?
6 MR. SILADY: If there was a catastrophic failure 7 of the vessel most lip.ely it would damage the surrounding 8 area and you would end up being depressurized.
9 I am going to show you the results of the heat to removal when you are in depressurized condition. With 11 regards to heat removal, you can remove the heat out to 12 the ground even if you don't have the reactor cavity 13 cooling system cnce you are depressurized. But for a
- ' (:) 14 catastrophic vessel failure you may have other concerns.
15 (Slide.)
4 16 This is the same graph, and in the interest of 17 time, the only thing that is different on it is that the la natural convection now with the helium being depressurized 19 is negligible and all the heat is removed radially by 20 conduction of radiation.
21 (Slide.)
22 These are the temperatures with the reactor 23 cavity cooling system removing the heat in the pressurized 24 and depressurized case.
- 25 The normal operating temperature, average 4
--r ,-,---g ,-r- - - , , -~--,--,-,,,.~n--.-.a, -n
<,,y,-- .,m7 ,,,.y--,--m-,re,--,,w-m,-- - , - p,, ,,-,,,y,--w,- - - - - -
,,v- -~,,y,,,, , ..r,-
212 L 1 temperature is approximately 650 degrees C. ~ This should 2 start a little bit lower. And the average then never 3 exceeds what our normal operating peak design limit is,
! 4 which is 1250 degrees C, nor does the maximum.
5 So we stay in the-pressurized condition below.
f 6 the peak temperature for which the fuel is designed.
7 In the case ~of the depressurized transient,
- .a however, we do go over the normal operation design limit.
- 9 On average we go just 50 degrees or so'above it, but in l 10 the middle of the core the temperature reaches
! 11 temperatures on the order of 16 to 17 hundred degrees C.
12 (Slide.)
l 4-l 13 -Now the sizing of the core and the geometry and i f 4-}' 14 the power density were all determined so that that
- 15 temperature was at that point because we have extensive I 16 data that shows that the fission product, and particularly 17 that silicon carbide layer remains intact to very high 18 temperatures approach 2000 to 2100 degrees C.
l 19 Our first data-point that we pick up any l 20 appreciable release at all, and it is hardly measurable, 3
! 21 was in the range of 16 to 70 hundred degrees C. This is 22 data is for our reference fuel, the Triso coated UCO.
4 23 The normal operating-peak is approximately 1100, a
24 and.I said already that the design limit is 1250. What 25 back here at 650 is our average peak temperature. So i
Ill
1 213 O there is considerable margin here, and with the Am/ 1 4
2 characteristics that have been mentioned, the heat can be 3 ' removed.
4 (Slide.)
5 Taking a look at what that means in terms of 6 fission product release, since under the depressurized 7 case we do exceed the normal operating temperatures, there 4 8 are three sources for fission product release in the core.
, 9 One is the contamination, the heavy metal j
i 10 contamination where the UCO did not get on the inside of 11 all the coatings, the second one, which has already been 12 mentioned, is the defective particle as manufactured, and 13 the third source is any incremental failure that might O 14 occur during normal operation.
i 15 We calculate with these temperatures at i
16 considerable detail throughout the entire core and 17 throughout the 1000-hour transient the sum of the release la from each o,f these contributions.
19 We find then that the total release is 20 approximately 2 times 10 to the minus 5 which already 21 meets the 3 times 10 to the minus 5 needed in order to i
22 meet 10 CFR 100. But since it is coming out over hundreds 23 of hours, it is a very long-term release, and the reg.
I 24 guide meteorology would also probably buy you another 25 factor of 3 here.
!dit x yr--t ,, ,.c- -+ -- , ww --
g w-- - -- wr--+rr-,-----i. e v, .+em-, .yme-e-r, i-~,c. ...,-,-=.++r-+.-,n---,,--,
,,-,e---~%---m+-~-w-,
4 s 214 i
~~l t_,) 1 So all together, there-is margin in meeting 10 2 CFR 100 when you lose all cooling.
3 That concludes t'he remove the core heat
'4 ~ function.
5 Moving on to the second of those three 6 subfunctions ---
, 7 MR. REMICK: Question.
8 MR. SILADY: Yes.
9 MR. REMICK: What is the. vessel temperature in 10 the case of depressurized loss of cooling, loss of active 11 cooling and how high a temperature?
12 MR. SILADY: It approaches about 900 degrees F 13 when you are deprescurized. When you are pressurized it
~
i "
t 14 is more like 800.
4 i 15 (Slide.)
a 16 The design selections now ---
17 MR. NEYLAN: You might mention that that is some 18 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> into the transient.
I 19 MR. SILADY: Yes. In'both cases it is a very
. 20 long transient, hundreds of hours before the peak 21 temperatures are reached, 20 to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> in the core and 22 longer on the vessel, and the heat gets radiated out.
23 MR. SHEWMON: Would that be a ferritic or an i
24 austenitic vessel?
25 MR. SILADY: It is a low allow.'
J
_._I_________._____.___._-__._________________._____
. ..~ . .- . -- .- - - - - ._. .. .-
235
)
~
MR. NEYLAN: Ferritic.
)-
2 VOICE: 'PWR.and it would be LWR material.
3 MR. SILADY: Yes, standard.
4 MR. CARBON: Let me say for the committee, that ,
5 was the case of losing all the pumps and perhaps pipe
~
i 6 break depressurization, but with the control going in. ,
7 MR. SILADY: Yes, that is with-it scrammed, and 8 coming up in.a minute I am going to show it-without it ;
9 being scrammed.
10 First, I want to talk about controlling chemical i
11 attack first by water.
12 (Slide.)
13 We are very sensitive to graphite oxidation both 14 from the point of view of water and air, and so I 15 specifically wanted to get through those two.
16 In this design we have limited t he sources. We
- 17 have already talked about the tact that we have got i
- la magnetic bearings. The high-quality fuel is very f 19 important because the fission product release, it is all i 20 inside the fuel particles, and the only way the water can
- r 21 affect them is if the fuel particles failed.
4
{ 22 If the particles are intact, it-will not affect-
}; 23 the fission product release. When it is 'dled it can l 24 hydrolyze the fuel, the carbide fuel.
j 25 We also have a very reliable detection isolation t
i
, .. ,,.,,..-.-~,-,-,,%,-,_.--..m.,-,--~---,m . . . . . . . - m_..
~ ,,.-, - , ..,,,,------y-,v..,m,,-,m_,-,,,,,,.___,~._..
216 r'3
't_; I system similar to the one at Fort St. Vrain, except now we -
2 only have a single loop as opposed to some of our other i 3 former designs and this will make it even better in being 4 able in not needing to identify which loop it is coming 5 frem.
6 (Slide.)
7 Now again the fuel quality is the key.
j 8 Hydrolysis is misspelled a coaple of places on the graph, 9 but only those particles that have failed coatings are 10 candidates -- or that amount of contamination that you 11 have, can there be any release from those particles as a 12 result of water ingress, and that is a very small 13 fraction.
j 1
O 14 In addition, only that fuel within the particle 15 that is a carbide will hydrolyze in the period of time we i 16 are talking about. So there is even less carbide than 17 there is oxygen. So there is another factor here.
18 Combining the two, once again, for any anount of j 19 water coming in, unlimited amount of water, we can show 20 that we meet the construction permit leve] of 10 CFR 100.
21 We can hydrolyze all the fuel that is available to be 22 hydrolyzed and still meet 10 CFR 100.
, 23 (Slide.)
24 Now to control chemical attack by air, we have 25 the following design selections.
)
. _ _ _ . . _ _ _ _ _ . , , . _ . . _ , _ _ ..c...,_.___..._ - . . . , _ . . _ _ _ . - _ , , . . _ . . . . . . _ , . . , _ . . . - . . . . . _ . ~ . . . . . . . , . , . . _ . , . _
217
.~
1 First of all, of course, helium is non-reacting 2 and it is pressurized and the helium would have to go out 3 before you might get any air in. We have a very high 4 quality vessel, as already noted, in terms of the LWR 5 standard.
6 The fuel itself, and probably the next key point 7 is that the fuel itself is embedded in that matrix in the 8 block. So any air that might come in is first going to 9 react with the reflector or with the graphite around the 10 coole.nt channel. The air has to reach the particle before 11 there would be any oxidation of the particle.
12 (Slide.)
-s 13 This graph shows the results for two size 14 leaks. Of course the smaller leak is the much more 15 Probable case, and it takes a considerable amount of time 16 for the helium to blow down before the air through the 17 hydrostatic displacement and through the contraction of 18 the remaining helium would cause more air to come in.
19 When you get to larger leak sizes, for instance 20 this is comparable to approximately a four-inch relief 21 valve line, if you had a spurious relief valve lifting and 22 a failure open, the air can get in much more quickly.
23 The key probably is not so much the time cxis, 24 but the ordinate here in terms of the fraction of the core 25 graphite reacted, less than 100th of a percent for either 8
1
218 7- .q
(, 1 one of these cases, and as an additional bound if you take 2 all the air in the reactor building surrounding the 3 vessel, you are only at 10 to the minus 2. You can react 4 all of the air and you have only oxidized one percent of 5 the graphite in the core.
6 MR. SHEWMON: You wall is built to maintain 7 that, or does it generate more gas than it forms?
8 MR. SILADY: The wall? Excuse me. I didn't 9 understand the question.
10 MR. SHEWMON: The wall which is air tight so you 11 don't have more oxygen coming in from the outside.
12 MR. SILADY: It is true that the reactor
._ 13 building is not a conventional light-water reactor, 14 pressure retaining containment. It is a standard 15 structure that has vents that open up and then close, 16 depending upon what the pressure is inside the building.
17 Those vents, however, are at a lower elevat-lon than the 18 reactor vessel. There could be some additional air coming 19 in, but it would be quite small.
20 I give the number of one percent sort of as a 21 bound that somehow you could get all that air in through 22 one of these leaks.
23 MR. NEYLAN: It also has to displace the helium.
24 MR. SILADY: Yes, that is another point of 25 course. As you blow down through sny of these leaks, the 8
219 k[ 1 helium is going to displace the air and push the air out.
2 MR. MARK: Those numbers of amounts of the fuel 3 affected are calculations or are measurements?
4 MR. SILADY: These are calculations in terms of 5 the oxidation. With regards to the heat removal, there 6 are data from AVR and there is data from intentional 7 shutdowns of poor circulation at Fort St. Vrain that 8 verify the conduction and the heat capacity and the heat 9 removal.
10 MR. CARBON: Mr. Silady, could you wind up now.
11 MR. SILADY: Okay, very good.
12 (Slide.)
,, 13 This is the passive retention air ingress, and i
la it is much the same as the others.
15 Let me move on to the reactivity.
16 (Slide.)
17 The design selections for reactivity are the la negative chemistry coefficient of the fuel, the control 19 rod system and the reservo shutdown system.
20 In this case though none of those control i
21 reactivity design selections come into play with the l l
22 exception of the fuel and the negative temperature 23 coefficient.
24 The LEU fuel is very negative, and it is on the l 25 order of 4 to 6 times 10 to the minus 5. So one sees that i 1
8
220 F7
_j 1 in the pressurized conduction cooldown that there is very 2 little change in the average or in the maximum temperature 3 with or without trip. So this is no inward rod motion and 4 this is a complete loss of flow pressurized. There is a 5 similar situation depressurized.
6 MR. SHEWMON: The temperature coefficient is 7 what shuts it down then?
8 MR. SILADY: Yes, it is, and that is why the 9 transients are show out to 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. After 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> your to xenon has decayed and at that point you would want to 11 either insert another one of your control systems, either 12 the control rods or the reserve shutdown system, or
_, 13 possibly, and we are looking into it, maybe the negative i
(~~
14 temperature coefficient is such that these temperatures 15 will not ride up even to very high temperatures.
16 There was one thing I should mention. This has 17 been run at AVR in Germany. They locked out their control la rods and they shut off their circulation and they saw this 19 behavior where the heat could be removed, the negative 20 temperature coefficient would turn it around and they 21 would not damage any of their fuel.
22 In their case it wasn't 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, but it was 23 23 hours. When they went critical again they saw they had 24 generated a small amount of power, whatever the cooling 25 system was removing.
8
221 1 Yes.
2 MR. SIESS: At which reactor was that now?
i 3 MR. SILADY: That was AVR. It is approximately l '4 45 or 46 megawatts.
) 5 (Slide.)
6 In conclusion, high~ quality fuel limits the 7 normal operation radionuclide inventories in the primary i
8 system to levels that are within 10 CFR 100. The 10 CFR 9 100 limits can be met for all events relying on only 1
1 to passive design selections to retain the fission products.
?
11 And although I didn't stress it, but it is in
- 12 your package, the protective action guide limits can be ,
i 13 met as well.
j 14 (Slide.)
15 In summary for all three presentations, the 16 inherent characteristics and passive features assure the 1
17 radionuclide retention in the fuel sufficient to obviate 18 the need for offsite sheltering or evacuation or emergency
) 19 drills.
l l
20 The program utilizes a systematic traceable
} 21 licensing approach that we didn't get a chanced to go into i
22 today, but which is specific to the modular HTGR that is 23 consistent with the advanced reactor policy.
l 24 Any additional questions?
25 MR. SHEWMON: The last bullet you have there, t
i r 4
k 5
222
. , I presumably this is an improvement over what somebody else 2 has done or you wouldn't say anything about it. Does that 3 come from the fact that you do in a factory instead of 4 being field erected or what does it mean?
5 MR. SILADY: You are on the last ---
6 MR. SHEWMON: The program utilizes systematic, 7 traceable licensing approach specific to modular HTGR.
8 MR. SILADY: Yes. In the first presentation you 9 will find in your package a summary of our licensing 10 approach. Mr. Millunzi didn't get an opportunity to go 11 into, but it utilizes risk assessment to choose the design 12 basis events, it utilizes the functional analysis in the
,_ , , 13 risk assessment to come up with principal design criteria 14 and which systems and structures should be safety related.
15 MR. NEYLAN: I think the question asked about 16 the construction of the factory assemblies. This is a 17 system engineering approach to the design which is 18 independent of the concept in reality, but it is something 19 that we are developing under the DOE program.
20 MR. SILADY: Yes, that is a more direct answer.
l 21 MR. MARK: Can you simply explain why the 22 temperature runs a hair higher without trip than it does 23 with trip?
24 MR. SILADY: Yes. It takes a finite amount of 25 time for the highly negative temperature coefficient to 8
223 7
1 turn the power generated around, and that delta then comes 2 in in the first seconds. It is something like a minute or 3 two and we are back down to the decay heat levels. But in 4 that period that amount of energy was added to the system 5 and that is why the temperature from that point forward 6 runs a little bit higher. It tends to narrow with' time, 7 and at long times it would look perceptibly different.
8 MR. CARBON: We are running out of time.
9 Bill, did you have a question?
10 MR. KERR: You have convinced yourself that it 11 is impossible for a graphite to burn?
12 MR. SILADY: Yes, we have. We have taken a look 13 at a number of wild scenarios and I am convinced.
i
's 14 MR. REMICK: If the emperor removed all 15 requirements for emergency planning, is there anything you 16 would do different, or maybe the inverse of that, are you 17 doing anything specifically so you don't have to worry la about emergency planning?
19 MR. SILADY: Specifically we are putting our 20 emphasis on fuel and on the passive characteristics.
21 MR. REMICK: But would you be doing that anyhow 22 or not?
23 MR. SILADY: I think that we have been driven by 2a that requirement from our owners.
25 MR. CARBON: Fine. Thank you, Mr. Silady.
8
224 i
F7 l
.s 1 Thank you, gentlemen. Very good.
2 Mr. Chairman, I return it to you then.
3 MR. WARD: Thank you very much, Max.
4 (Whereupon, at 6:30 p.m., the recorded 5 proceedings of the 311th ACRS General Meeting concluded.)
6 * * * *
- 7 8
9 10 11 12
- , 3 13
~- ja 15 16 17 18 l
19 l
20 21 22 23 24 25 l
CERTIFICATE OF OFFICIAL REPORTER O
This is to certify that the attached proceedings before the UNITED STATES NUCLEAR REGULATORY COMMISSION in the matter of:
NAME OF PROCEEDING: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 311TH GENERAL MEETING i
DOCKET NO.:
1 PLACE: WASHINGTON, D. C.
DATE: FRIDAY, MARCH 14, 1986 ,
i were held as herein appears, and that this is the original transcript thereof for the file of'the United States Nuclear Regulatory Commission. .'%
l 4
(sigt) zu rt l
(TYPED) 4 DAVID L. HOFFMAN Official Reporter ACE-PEDQRAL REPORTERS INC.
Reporter s Affiliation, O
CERTIFICATE OF OFFICIAL REPORTER 6
This is to certify that the attached proceedings before the UNITED STATES NUCLEAR REGULATORY COMMISSION in the
, matter of:
! NAME OF PROCEEDING: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 311TH GENERAL MEETING l
i 1
1 DOCKET NO.:
PLACE: WASHINGTON, D. C.
1 O DATE. rR1DxY, nARCu 14, 198e were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission.
e (sigt h M j i
^
(TYPED) ;
MARY C. SIMONS Official Reporter ACE-FEDERAL REPORTERS, INC.
'> Reporter's Affiliation O ,
O NRR STAFF PRESENTATION TO THE ACRS ,
l
SUBJECT:
IMPLEMENTATION PLAN FOR THE SEVERE ACCIDENT POLICY STATEMENT AND THE REGULATORY USE OF NEW SOURCE-TEP,M INFORMATION
, DATE: MARCH 14, 1986 l CaeSENTER: z0LTAN R. ROSzTOCzY 1
l PRESENTER'S TITLE / BRANCH /DIV: CHIEF i
REGULATORY IMPROVEMENTS BRANCH l DIVISION OF SAFETY REVIEW & OVERSIGHT l
l PRESENTER'S NRC TEL. NO.: 492-8016 l
1 l
1 l
O
, (y v
p v
s -
SEVERE ACCIDENT POLICY STATEMENT - ACTION ITEMS POLICY STATEMENT
~
NEW APPLICATIONS EXISTING PLANTS sm GUIDANCE ON THE ROLE SYSTEMATIC APPROACH OF PRAs FOR THE EX MINATION OF INDIVIDUAL PLANTS PERFORMANCE CRITERIA IMPLEMENT MODIFICATION FOR CONTAINMENT THROUGH BACKFIT POLICY SYSTEMS CHANGES IN RULES AND REGULATORY PRACTICES, '
AS NEEDED i,; .
hl o, _ _ _ _ _
() IMPLEMENTATION PROGPAM El.EMENTS
- 1. EXISTING PLANT EXAMINATION
- REVIEW OF THE IDCOR INDIVIDUAL PLANT EXAMINATION METHODOLOGY
- DEVELOPMENT OF GUIDELINES AND CRITERIA FOR PLANT EXAMINATIONS
- 2. DEVELOPMENT OF GUIDANCE ON THE ROLE OF PRAs FOR NEW PLANT APPLICATIONS
() - DETERMINISTIC REQUIREMENTS
- ACCEPTABLE CONTENT OF PRAs
- CPITERIA FOR THE REGULATORY REVIEW AND INTERPRETATION OF THE PRA RESULTS E3E gy
- 3. CHANGES IN RULES AND REGULATORY PPACTICE
- SOURCE TERM PELATED CHANGES
- SEVERE ACCIDENT RELATED CHANGES O
2
l II) h j s I
EXPECTED ACCOMPLISHMENTS i
- PLANT SPECIFIC VULNERABILITIES WILL BE IDENTIFIED AND'
! FIXED (BACKFIT RULE) i i IF GENERIC VULNERABILITIES ARE IDENTIFIED, APPROPRIATE DESIGN AND/0R OPERATIONAL CHANGES WILL BE REQUIRED (RULEMAkING) l
([)
- LESSONS LEARNED WILL HELP DEVELOPMENT OF IMPROVED DESIGNS WITH SAFETY BENEFITS
, A NEW, MORE REALISTIC REGULATQBX. APPROACH ON SOURCE TERMS
. WILL BE PURSUED (SOURCE TERM R'kfATED (HANGES) i i
j . i i
i l
1
! () .
l 3
f i V
SUMMARY
U OF EXPECTED ACCOMPLISHMENTS SEVERE ACCIDENT POLICY FMPLEMENTATION 6/86 COMPLETE THE NRC ANALYSIS OF SIX PEFERENCE PLANTS FOR SEVERE ACCIDENTS INCLUDING SOURCE TERM CALCULATIONS RESOLVE IDCOR/NRC TECHNICAL ISSUES 7/86 COMPLETE THE REFERENCE PLANT SENSITIVITY 7/86 STUDIES (EVALUATION OF UNCERTAINTIES)
COMPLETE REVIEW 0F IDCOR METHODOLOGY 10/86 FOR INDIVIDUAL PLANT EXAMINATIONS U 12/86 BRIEF COMMISSION ON THE FINDINGS AND PECCFMENDATIONS FOR THE INDIVIDilAL PLANT EXAMINATIONS ISSUES GUIDANCE RQg PUBLIC COMMENT 2/87 ON THE ROLE OF PRAs FOR'NEW PLANT APPLICATIONS ISSUE FOR PUBLIC COMMENT RULE CHANGES 4/87 NECESSARY TO PESOLVE GENERIC SEVERE ACCIDENT RELATED VULNERABILITIES 1
()
v 4
O O O -
/
Reference Plant G /
\
- I Analyses \ l Regulatory }
\ 6/86 Principle j
/ \
N j g 4/86 j
- - - -- (;..
Preparation of Strawman
' [ Guidelines 9/86 II
, - - ,\
Technical issue Evaluation of I.'nal /
Guidehnes Research Resolution + Reference Plants L 60 Is Critena l Update I 7/86 \ 10/86 /
10/86 g j
~- ---
! Development of Proposed j(
Criteria 9/86 tn / Ser:sitivity \ Coramission I Analyses I Briefing
\ 7/86 / 12/86
/
w lf Standards for Review of Acceptable IDCOR Methodology Methodology 3/86 10/86 LEGEND Program Activities
'~ C -~~~'N 4-
) Inp.Jt from other Programs \
- / IDCOR l Methodology I
\ 7 //86 /
i Figure 3.1 .
Program Element 1 - Development of Guidance for Individual Plant Examinations i1 i
p O l
TABLE 3.2 j
LISTING OF MILESTONES ,
3.1 REVIEW 0F THE IDCOR INDIVIDUAL PLANT EXAMINATION FETHODOLOGY STANDARDS FOR AN ACCEPTABLE METHODOLOGY 3/96 SUBMITTAL OF IDCOR REPOPTS FOR Tk0 EWRS 3/86
- AND Th0 PWRt
(
SUBMITTAL OF REMAINING THREE IDCOR REPORTS 7/86 l0 NRC CCn +vrs T0 iDC0a ON Metu0DOLOGY -
7/86 I -
EVALUATION OF THE APPLICATION OF THE IDCOR 10/86 i
E TH000 LOGY TO SEVEN PLANTS I
l 3.2 DEVELOPWlif 0F GUIDLINES AND CRITERIA FOR PLAffT EXAMINATIONS i i
i 3.2.1 TECHNICAL ISSUE RESOLIITION i
I - DEFINE UNCERTAlflTY PANGES FOR SURRY, -
2/86 !
1 l PEACH POTTOM 8 SEQUOYAH l
- DEFINE UNCERTAltfiY RANGES FOR ZION, -
3/86
! GRAND GULF 8 LASALLE O -
DRAFI NRC/IDCOR ISSUE PAPERS -
5/86 1
FINAL NRC/IDCOR ISSUE PAPERS -
7/86 i
6 l
TABLE 3.2 (CONTINUED) 3.2.2 EVALUATION OF REFEPENCE PLATHS COMPLETION OF IDCOR SEVERE ACCIDENT EVALUATIONS (WITH UNCERTAINTY ANALYSIS) - 7/86 COMPLETION OF SARRP RISK EVALUATIONS:
SURRY -
4/86 '
PEACH BOTTOM -
5/86 ZION -
5/86 SEQUOYAH -
6/86 GRAND GULF -
6/86 Q -
CC W LETE PEFERENCE PLANT RISK PROFILE -
8/86 l EVALUATION OF REFERENCE PLANTS -
8/86 3.2.3 PPEPARATION OF STRAWAN GUIDELINES 3.2.4 DEVELOPEffT OF PROPOSED CRITERIA STRAWAN GUIDELINES & PROPOSED CRilERIA PEACH E0110M -
6/86 OTHER PLANTS -
9/86 3.2.5 DEVELOPENT OF FINAL GUIDELINES & CRITERIA O -
FINAL GUIDELINES 8 CRITERIA -
10/86 s- 7
~
O O O -
/ \
l Resolution of USis and GSis g
3
\
/~__-_,\ \ _____/ Evaluation of Deterministic f Assessment of 1\ Deterministic -
Requirements Reference Plants - Requirements "
12/86
{ GESSAR & IP - "
10/86 y~ sin J
I
Procedures k Guides 9/96 Procedures for
\ s- / Core Damage r
--.--.s ____- rreou.nc,
\ 10/86
/ IDCoR l Methodology G "* "
Commission Commission
( 10/86 l
nim m -
- Paper - Approval w_-, Content,of gg gg PRAs 1/87 2/81
/ r - ~ ~ ~~~ Procedures for 00 / Guidelines and Containment b g Criteria for IPE Consequence ,
10/96 / Analyses
\ / 11/86
/, - - ,\
[ Reference Plant Analysis
\ 6/86 / Use of Safety
\ %---- '
/
'_ ,N Goals With Uncertainties
/ Safety Goal b \ S/86 Containment l Performance Guidance on
\ Objective / Criteria for 6/96 N%---s/ Regulatory Use 12/86 LEGEND f'___ % Criteria for incremental Risk
/ PSA Manual. \ l l Program Activities '
g PR A [nsights S/86 a Reports f (,,,,_,) input from Other Programs
\ 9/85 j
% - _. s Figure 4.1 Program Element 2. Development of Guidance on the Role of PRAs l
I t
O "%
(_
[')N
- Source Term Calculation \
\
for Reference Plants l
\ _ - 6/86- - -
/ Development of \ Capability for Development of I Source Term Codes I '
Source Term Calculation '
New Source Terms - Source Term
\ 12/85 / 3/86 12/8G r Related Changes
\ ~--__-s /
Selection of /
Regulatory Principle Research Update 4/86 10/86 l
\---- _ _ _ /
N g/'"~ Containment g Development of to Performance - '
Containment
\ Design Objective ' '
Performance Criteria
\ 6/86 / 10/86
~ - - - - - - - '
/ Assessment of \
Reference Plants Containment Related l Changes g 6/8G j
identification of Generic Vulnerabilities 9/86 Other Severe Accident Related Changes p------% p--- -
[ Resolution of [ Examination of LEGEND ) !
( USis and GSis j { individual Plants l l Program Activities ( ._ _,, ___/ \ ,, _ _ _ ,, , _ s
, _ . , ~
( ) Input from Other Programs Figure 5.1 Program Element 3 - Changes in Rules and Regulatory Practice j :
h
4 O
l POIENTIAL SOURCE TERM CHANGES NEAR-TERM INTERMEDIATE LONG-TERM
! REVISED TREATMENT EMERGENCY PLANNING SITING 0F ACCIDENTS IN EIS i
l y
CONTAINMENT LEAK RATES ACCIDENT
- ['mREMOVALOFSPRAY ADDITIVES (PWR) AND INTEGRITY MONITORING SUPPRESSION P0OL ENV QUALIFICATION CPEDIT (BWR) 0F EQUIPMENT e-SAFETY ISSUE EVALUATION i l
l 4
i CONTROL ROOM HABITABILITY i 4-AND AIR FILTRATION SYSTEMS O
,. . - - - . ._ 10 _ _ _ _ _ _ . _ . _ ._ ._.- i
O POSSIBLE REVISED APPROACH i
WISH TO RE-EXAMINE THE USE OF FISSION PRODUCT CALCULATIONS IN LICENSING (TID-14844 AND REG.
GUIDES 1.3/1.4) ,
POSSIBLE APPROACH 1
CONSIDER A NUMBEP 0F ACCIDENT SEQUENCES LEADING TO j COPE DEGRADATION, MELT AND RELEASE INTO CONTAINMENT
()
- EVALUATE ACTIONS OF NATURAL REMOVAL PROCESSES AS WELL AS FISSION PRODUCT CLEANUP SYSTEMS TO EVALUATE PROGRESSION OF' ACCIDENT AND CALCULATE RELEASE OF FISSION PRODUCTS TO CONTAINMENT FROM ABOVE, DETERMINE FISSION PRODUCT TYPES, AMOUNTS AND TIME-DEPENDENT CONCENTRATIONS IN CONTAINMENT THAT GENERALLY ENVELOPE SEQUENCES ESTIMATE TIME-DEPENDENT CONTAINMENT LEAK RATE AND CALCULATE FISSION PRODUCT LEAKAGE FROM CONTAINMENT CALCULATE DOSES TO HYPOTHETICAL INDIVIDUALS AT EAB
([) AND LPZ AND COMPARE TO PART 100 (MAY HAVE TO ADD OTHER ORGAN DOSE CRITERIA)
-___--------_--__-_-----_--_-------_--_liL--__------_-----_-_---
SUMMARY
OF EXPECTED ACCOMPLISHMENTS SOURCE TERM RELATED CHANGES 4
ISSUE FOR COMMENT REVISED SRP SECTION 6.5.2 9/86 i SPECIFYING THE NEED FOR SPPAY ADDITIVES IN PWRs ISSUE FOR COMMENT REGULATORY GUIDE 1.3 AND THE , 9/86 APPROPPIATE SECTION OF THE SRP ON FISSION PRODUCT
., SCRUBBING IN SUPPRESSION POOLS (BWRs)
ISSUE FOR COMMENT PROPOSED CHANGES TO 10 CFR 50.47 2/87 AND 10 CFR 50, APPENDIX E ON EMERGENCY PLANNING REVISE NRR OFFICE LETTER 16 WITH RESPECT TO THE 2/87
]
I USE OF SOURCE TERMS IN SAFETY ISSUE EVALUATION .
- O I ISSUE FOR COMMENT CHANGES IN CONTAINMENT LEAK 3/87 I RATE REQUIREMENTS, INCLUDING POTENTIAL CHANGES l IN 10 CFR 50 APPENDIX J REVISE 10 CFR 50.49 AND REGULATORY GUIDE 1,89 6/87 WITH RESPECT TO THE RADIATION ENVIRONMENT FOR EQUIPMENT QUALIFICATION, FOR COMMENT BY ISSUE FOR COMMENT REVISIONS OF SITING CRITERIA 10/87 (10 CFR 100) BASED ON NEW SOURCE TERM INFORMATION ISSUE FOR COMMENT REVISED REGULATORY GUIDE 1.97 12/87 ON ACCIDENT MONITORING AND MANAGEMENT ,
O 12
O RetAT10sSsleS wlTs OTnER eR0 GRAMS RES PROGRAMS ,
- NUREG-0900 SUPPLEMENT, RESEARCH PLAN FOR SEVERE ACCIDENTS
- NUREG-0956, REASSESSMENT OF THE TECHNICAL
! BASES FOR ESTIMATING SOURCE TERMS '
- NUREG-1150, PEFERENCE PLANT ASSESSMENT
- UPDATE ON SEVERE ACCIDENT RESEARCH
! NRR PROGRAMS 1
- SAFETY GOALS: FINAL VERSION, CONTAINMENT
]
PERFORMANCE DESIGN OBJECTIVE O - UNPESOLVED AND GENERIC SAFETY ISSUES:
STATION BLACKOUT, SHUTDOWN DECAY HEAT REMOVAL
- PPA REVIEWS AND INSIGHTS PEPORTS: INDIAN POINT, ZION, LIMERICK, AND GESSAP; PPA INSIGHTS REPORTS, PROCEDURES GUIDE AND REVIEW MANUAL INDUSTRY PROGRAMS ,
- IDCOR: REFERENCE PLANT ANALYSES, TECHNICAL ISSUES, INDIVIDUAL PLANT EXAMINATION METHODOLOGY
- AIF: SOURCE TERM ISSUES, PRA ISSUES
,O 13
(] CONSIDFPATIW OF EXTERNAL EVENTS IN THE IFPLEltNTAT10,NPROGRAM IDCOR WOPK, llP TO NOW, AND NRC WORK IN SUPPORT OF NUREG-1150 WEPE LIMITED TO INTERNAL EVENTS, IN GENERAL.
SEISMIC EVENTS WERE CONSIDEPED UNDER SEPAPATE PP0GPAMS, NPC HAS A SEPAPATE SEISMIC PROGPAM, INDllSTRIES EFFORTS ARE C00PDINATED BY THE SEISMIC OUALIFICATION USERS GROUP (SQUG), SENIOR SEISMIC REVIEW ADVIS0PY PANEL, EPRI, AIF ETC.
O DIE G0AL OF THE POLICY STATEFENT AND ITS IMPLEfflfTATION IS TO BRING STABILITY TO LICENSING AND PEGULATION WITH RESPECT TO ALL SEVEPE ACCIDENT ISSUES. THIS CANNOT BE ACCOMPLISHED WI$HOUT CONSIDEPATION OF EXTERNAL EVEffTS.
THE AVAILABILITY OF ANALYTICAL METHODS FOR EXTERNAL EVENTS IS COMPARABLE TO THAT OF INTEPNAL EVENTS. BOTH NRC AND THE ItRISTRY HAS DEVELOPED METHODOLOGY FOR SEISMIC RISK ANALYSIS.
APPROXIMATELY 20 SEISMIC PRAs HAVE BEEN COMPLETED OR ARE CLOSE O
II4
/ '3 TO COMPLETICN. SIPPLIFIED METHODS APPROPRIATE FOR IDENTI-L.)
FYING VULNEPABILITIES AFE BEING DEVELOPED BY BOTH NPC AND EPRI. SAMPLE APPLICATIONS OF THE PETHODS APE PLANNED.
AN APPROACH TO THE HANDLING OF EXTERNAL EVENTS IN THE IMPLE-FENTATION PROGPAM HAS BEEN D$VELOPED BY THE NRC STAFF AND ,
IT IS UNDER MANAGEPENT PEVI&l.
n V
i e
=
\
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15 ,
RECterENDED APPROACH TO EXTEPNAL EVENTS O
EXTERNAL EVENTS NEED TO BE CONSIDERED; HOWEVER, THE EXIENT OF THE REVIEW AND THE SCHEDULE FOR>EXTEPNAL EVENTS DOES NOT HAVE TO BE THE SAME AS FOR INTERNAL EVENTS.
~
THE EFFORT SHOULD CONCENTRATE ON FLANI VULNERABILITIES DUE TO EXTERNAL EVENTS. RESOLUTION OF THE QUESTION WHAT IS THE CON-TRIBUTION OF EXTERNAL EVENTS TO OVERALL RISK IS NOT A NECESSITY, THE FIPST PHASE OF THE PROGRAM SHOULD CONCENTRATE ON:
A.) ESTABLISHING THE EXTENT TO WHICH PLANTS HAVE. BEEN REVIEWED FOR EXTERNAL VULNERABILITIES IN THE PASI OR WILL BE REVIEWED UNDER ONGOING PROGRFS, LIKE A-46.
B.) ESTIMATING THE MARGIN THESE REVIEWS WILL ASSURE PELATIVE TO EVENTS BEYOND THE DESIGN BASIS, FOR EXAMPLE EARIHOUAKES BEYOND SSE.
C.) IDENTIFYING EXTERNAL EVENTS THAT NEED-TO BE INCLUDED IN THE VULNERABILITY SEARCH.
D.) IDENTIFYING THOSE AREAS WHERE EXAMINATION FOR VULNERABILITIES IS NEEDED, AND ARE NOT COVERED UNDER ANY UF IHE EXISTING O eROGRAnS.
i 16
RECTSDED APPROACH TO EXTEPNAL EVENTS " (CONTINUED) t .
a NRC SHOULD SOLICITY INDUSTRY PAPTICIPATION IN THE FIPST PHASE OF THE PROGRAM S
kHEN THE FIRST PHASE IS COMPLETE, THE STAFF WILL ADDRESS THE
- FOLLOWING OUESTIONS
1
~
i A.) IS THEPE A NEED TO CONTINUE THE PROGRAM?
B.) kHAT FORM SHOULD THE SECOND PHASE OF THE PROGRAM TAKE?
]
- O i
J I
4
- O ,
t.
^
17
O NRR STAFF PRESENTATION TO THE ^
SUBJECT:
NRC ADVANCED REACTOR PROGRAM l
l DATE: mRCH 14, 1386 PRESENTER: THOMS L. KING l
)
l l
PRESENTER'S TITLE / BRANCH /DIV: SECTION LEADER l SAFETY PROGRAM EVALL'ATION BRANCH DIVISION OF SARTY REVIEW & OERSIGlT PRESENTER'S NRC TEL. No.: 49 2-7014 SUBCOMMITTEE: FULL C0ffilTTEE e
O
1 PROGRAM
()
I) ORIGINAL PLANS: .
- NRC REVIEW OVER THE NEXT TWO YEARS CONCEPTUAL DESIGNS FOR l ONE HTGR AND TWO LMRs.
- INTERACTIONS AMONG THE STAFF /ACRS/D0E/ REACTOR DESIGNERS ON KEY ISSUES FOLLOWED BY NRC REVIEW OF A PRELIMINARY
- SAFETY INFORMATION DOCUMENT (PSID) AND PRA ON EACH CONCEPT.
- REQUIRED STAFF RESOURC~S WERE 5-6 STAFF YEARS PER YEAR AND APPR0XIMATELY $1.25M TECHNICAL ASSISTANCE PER YEAR.
! ALL RESOURCES WERE TO BE FROM NRR.
~
II) CURRENT DIRECTION:
($)
- REDUCE EFFORT TO 2 STAFF YEARS PER YEAR AND NO TECHNICAL
- ASSISTANCE.
l
- IMPACT ON PLANNED REVIEWS IS CURRENTLY BEING ASSESSED.
WILL PROBABLY LIMIT FUTURE INTERACTIONS TO A FEW KEY ISSUES.
e i
7 4~
KEY ISSUES-EXAMPLES O
i i
i l 1) CONTAINMENT / CONFINEMENT ;
l 2). USE OF INTERFACE CRITERIA ONLY FOR THE BALANCE OF PLANT.
l
- 3) USE OF A SINGLE, PASSIVE, HIGHLY RELIABLE DECAY HEAT REMOVAL SYSTEM, l.
-l
. 4) TREATMENT OF SEVERE ACCIDENTS AND EMERGENCY PLANNING. ,
4 1
2 i
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i O
d
-1 l
i I
s I.
i i
- O l 1;
l l
.l
MAJOR ACCOMPLISHMENTS TO DATE
)
1 I) MEETINGS HELD ON KEY ISSUES:
- HTGR:
- TOP LEVEL DESIGN CRITERIA
- APPROACH FOR EMERGENCY PLANNING
,
- DESIGN APPROACH
--LMR:
- SHUTDOWN HEAT REMOVAL i -.
- LICENSE BY TEST APPROACH
- METAL FUEL
- APPLICATION OF LWR-GSIs g'-
- SAFEGUARDS AND SECURITY II) REVIEW:
- DOCUMENTS CURRENTLY UNDER REVIEW ON:
- LMR SHUTDOWN HEAT REMOVAL
- LMR SAFEGUARDS AND SECURITY
.2 I
,- .,-,,,,----,,<-,,-e,--,n,..n,-.a. ,,e - +.. , - - - - . .,,-r,- ,w -a- ,, ,,w,~~ .n e- we.---,,,- - -,. , , . , , ~ , , , - , , ,-.w,..,,,..-,.
h.
G A5 COOLID RE ACTOR ASSOCIATES
(~3 . . . . . . m ,:.,2 w 9 U :..w tA97.mi 6
... vom February 17, 1986 Honorable Don Fuqua Honorable James A. McClure Chairman, and Members Chairman, and Members House Committee on Science Senate Committee on Energy and
& Technology Natural Resources Honorable Tom Bevill Honorable Mark O. Hatfield Chairman, and Members House Chairman, and Members Senate Appropriations Subcommittee Appropriations Subcommittee on on Energy 6 Water Dev. Energy 6 Water Dev.
Honorable Morris K. Udall Honorable Alan R Simpson Chairman, and Members Chairman, and Members House Subcommittee on Senate Subcommittee on Energy and Environment Nuclear Regulation r-(%
~ ./
l
Dear Mr. Chairmen:
On behalf of the CCRA utilities and the maj or private-sector participants in the High Temperature Cas-Cooled Reactor (HTGR) Program, we respectfully submit a Joint Statement in support of the HTGR and sustained Government budget support for FY 1987. We look forward to the upcoming hearings and the occasion to expand upon the basis of our support for the HTGR.
Sincerely, R .% . uJ ask.r R. F. Valker, President Public Service Co. of Colorado Chairman, Gas-Cooled Reactor Associates Enclosure cc: Honorable John S. Herrington Secretary, Department of Energy Honorable Nunzio J. Palladino d Chairman, Nuclear Regulatory Commission
Jn!NT PARTICIPA' TS' SMTESf ENT M THE HTCR PROGRAM TO CONCRESSIONA1. OVEESIGHT CO?tMITTEES A secure, s c at. l e , and economic energy supply is vital to National security and
[h economic well-being. Other developed countries have recognized that nuclear power is an important resource in providing this necessity. In particular, other industrialized countries such as Japan, Germany and France, as well as emerging industrialited countries such as Taiwan and Korea, have undertaken ambitious, sustained nuclear power programs to assure their citizens and industries the advantage of reliable, inexpensive energy; to reduce the constraints of dependence on foreign sources for essential commodities; and to foster their nuclear industries' competitiveness in the international market. Also, the Soviet l'n io n and other industrialized socialist countries are continuing a methodical development and deployment of nuclear power for a broad range of energ applications.
The U.S. is struggling with how to revive its civilian nuclear power resources. 'Ji t hout presuming to offer a panacea, we point out that there is a clear need for stable licensing requirements and less onerous licensing procedures:
easily understood and acceptable risks to investors and the public; and plant sizes, invested capital, and construction schedules that are consistent with current load growth and financial realities.
The Modular High Temperature Gas-Cooled Reactor (tiHTGR) is well suited to fulfill these needs. Its core size and power density are limited by design such that the ceramic coatings of the fuel particles retain the fission products to an acceptable degree under all licensing basis events, with decay heat removal being accommodated by the passive mechanisms of radiation and conduction. This feature f will simplify the overall concept, particularly the licensing and operations (3) activities. In addition, safety related systems and components are minimized and are contained within a compact, standardized nuclear island. Major portions of the nuclear island can be shop f abricated and assembled to nuclear standards, facilitating enhanced quality assurance. The balance of plant is physically separated from the nuclear island and may be manufactured and constructed to conventional fossil plant standards. Inis will improve productivity, shorten the construction schedule and reduce field erection costs. The nuclear island is composed of independent reactor modules that can be constructed sequentially or as a clus te r to more closely match load growth requirements and the utilit,'s financing constraints. This feature will provide added flexibility and reduced financing risk that is most desirable in the current era of uncertainty for U.S . !
utilities' load growth forecasting. I l
The tiHTCR is also well suited to supply high quality steam, broadening its market through cogeneration or direct steam / heat supply. In addition, the MHTCE has the unique nuclear capability for advancing to closed-cycle, gas turbine and high temperature process heat applications. Further, the tiHTGR can be economicall.
deployed on a once-through, low-enriched uranium fuel cycle and converted to fuel recycle and/or high-enriched fuel cycles as economics, facilities and policies warrant. Considering the profound potential contributions to t;ational enerp security; the investeent to date and the relatively modest added investments that are required; the potential for international cooperation on the remaining development and demonstration as well as the broader international market '
potential, the Program participants submit that the continued development of th(
A MHTCR is worthy of fational commitment.
V At the present, the !!HTCR is undergoing its initial stage of design and licensing development. As a result of a thorough evaluation effort completed during FY 1985 within the DOE HTGR Program, the 350 M'J e prismatic annular core, MHTGR has been established as the reference design for U.S. development and l
T .
JOP:T PAPTICIPANTS' STATEh' T ON THE HTGR PROCPAM (CONT'D)
. _ near-term deployment. This concept makes maximum utilization of the HTCR ,
technology, design and demonstration experience established in the U.S. The w reference plant design adopts four reactor modules and two steam turbine generators for a nominal output o f 5 50 Mk'e . The evaluations performed to date indicate the plant design can m. t the performance, investment protection and econor -
requirements established through Gas Cooled Reactor Associates (GCRA), tne utility / user organization that supports the development and commercialization of the HTGR.
t Ongoing DOE Program activities should remain focused on the development of the reference design and key technology areas to support the preparation of a Preliminary Safety Information Document by the end of FY 1986 for formal review by the NRC in FY 1987. The goal of that review is the issuance of a licensability statement by the NRC for the MHTCR.
That will determine at an early stage whether the NRC accepts the proposed licensing criteria, the supporting safety analyses, and the MHTCR's technology development plan, including a demonstration project. ,
Those results will be essential to the private-sector participants in the development of their commitments to the demonstration and deployment of the MHTCR.
l As a complement to the DOE Program, GCRA, with cost sharing support from other industrial participants, has undertaken an initiative to develop a one module Demonstration Proj ec t . Through successful demonstration of the licensing, construction, performance, costs and schedule of such a plant as well as the associated development of an industrial infrastructure, commercial MHTGR plants will become a reality. Estimates to date indicate that such a Demonstration Proj e c t , including several years of operation, would cost approximately $70" million ('85 dollars) with an additional $115 million ('85 dollars) required for l J supporting technology development. k*ithout resource constraints, a Proj ec t could begin construction by 1990 and startup in 1994 Initiation of parallel commercial
) activity would permit commercial units to be deployed by the turn of the century.
j Ongoing Project development activities are focused on detailed demonstration requirements and plans, establishment of Project management and supply entities, and the establishment of a Project Strategy Plan that will address cost / risk sharing arrangements. It is recognited and accepted that the key to the success of
- the Proj ec t is the ability of the private-sector participants to raise a significant fraction of the Project funds.
4 Crucial to the furtherance of any U.S. MHTGR Project initiative is stable i
Government support of the reference plant design and licensing development as well j
as the associated technology development that supports both the Project and
! long-term higher temperature applications. Accordingly, it is respectfully recommended that Congress provide $32 million for continued MHTGR development i
within DOE's Advanced Reactor Program in FY 1987. In addition, it is recommended that $2 million be provided to the NRC for licensing review activities. L'h i l e this i
budget level will not support an aggressive Project schedule, it will sustain the i
existing resource base and communicate Program stability that will be invaluable
- for the continued development of the private-sector Project initiative and will foster the prospects for international cooperation.
~
This recommendation is submitted within the context of a balanced, affordable U.S. nuclear strategy that <
deploys improved light water reactor systems for the near-term and maintains a development program for a long-term nuclear fuel breeder.
l This Joint Statement is respectfully submitted to the Congressional Committees with oversight resp.onsibility for the Nation's nuclear power development programs i as evidence of the major participants' unified support for and recommendations on j the HTGR Pregram,
.-~'.-w - -- - - . , . - . _
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JOI':T PAoT? CI PA':TE' STATEME'i c': TuE HTCR PROCPAM /CO'T*Di N
\.
R.
- Y F. Valker, Chairman, President & CEO W M
' Harold U. Sonn, Chairman, President 6 CEO Public Service Company of Colorado Public Service Electric 6 Gas Company Chairman, Gas-Cooled Reactor Assoc.
/*
MsL. Everett, Chairman & CEO Ja E. Thomas. Ex:c. Vice President
'hiladelphia Electric Company 5. Diego Gas & Electric Company
. ice Chairman, Cas-Cooled Reacter Assoc.
E. A. Adomat, Executive Vice President Jaces E. Tribble, Pv6sident & CEO Florida Power & Light Co=pany Yankee Ato=ic Electric Company DJ/A Yo C. H. Dean, Chaircan, Board of Directors Berna /d C. Trueschler, Chairman & CEO Tennessee Valley Authority Baltimore Gas & Electric Company
$% a'a '
ni , o x > n$n1 J. DT G A t. Chairman, President & CE'O M th L. Turley, Chairman a d Pt.ilic Service Co:pany of New Mexico Arizona Public Service C piny C . m't A --
Girts Kru ins, President & CEO K Dance, President Co l o ra do - L't e Electric Association GA Technologies, Inc.
id *
' George /. Maneatis. Exec. Vice President
)- (
William C. Drotleff, Vice President Facilities & Electric Resources Devel. Stone & Webster Engineering Corp.
and Pres. Elect, Pacific Gas & Electric ftl.$'
Dobert J. d*Connor, President 6 CEO Donald M. Kerr. Sr. Vice President Idaho Power Cocpany EC6G, Inc. t s b A. JI fi ter, General Manager Do'nald E. i.yons ,/f re s ide nt f Salt Rt.er Project Power Systems C#up !
p Combustion Engineering, Inc.
V WW Leroy W. S tlair, President 6 C00 k
Howard Wahl, Vice President 6 Director New York Pvwer Authority Sechtel Power Corporation
T
+
l' NUREG 0313 REV. 2 EXPANDS REV. 1 COVERAGE INCLUDES ALL STAINLESS PIPING (CL 1, 2, 3)
REV. 1 HAD LIMITED SCOPE REQUIRES FORMAL QUALIFICATION OF NDE EXAMINERS AND PROCEDURES
. 7 46 s e n-L - wc a ch t2_ .
REV. 1 JUST RECOMMENDED THAT IMPROVED UT PROCEDURES BE USED O -
PROVIDES GUIDLINES FOR EVALUATION AND REPAIR OF CRACKED WELDS REV. 1 REQUIRED REPLACEMENT OF CRACKED WELDS 4
O !
r --
Q NUREG 0313 REV, 2 GENRALLY FOLLOWS RECOMMENDATIONS OF PIPING REVIEW _
COMMITTEE NUREG 1061 VOL. 1 RECOMMENDS:
USE OF IGSCC RESISTANT MATLS REPLACEMENT OF SUSCEPTIBLE PIPING PROCESSES FOR RESIDUAL STRESS IMPROVEMENT IMPROVED WATER CHEMISTRY PROVIDES SPECIFIC INSPECTION SCHEDULES CONSIDERING:
MATERIAL IGSCC RESISTANCE STRESS IMPROVEMENT PROCESSING WATER CHEMISTRY IMPROVEMENT REPAIRS AND CRACKING CONDITION PROVIDES GUIDELINES FOR CRACK EVALUATION AND REPAIRS UPGRADES LEAKAGE LIMITS AND MONITORING O
O 'O O 1
s PRESENTATION BY DOE AND ITS CONTRACTORS ON THE
'l MODULAR HTGR TO THE ADVISORY COMMITTEE i
i ON l
REACTOR SAFEGUARDS .!
MARCH 14, 1986 t
. _ . _ _ _ - . - , . _ _ _ - . _ _ . . . . . _ , _ _ _ . . . _ _ _ _ . , . . _ - , . . . . , _ ,..4
O O
[. -O i
1, ADVANCED REACTOR PROGRAM o l
i-o 0BJECTIVE n
DEVELOP A LOW COST, PASSIVELY SAFE, ELECTRICAL GENERATING OPTION COMPETITIVE h' -
WITH CONTEMPORARY ALTERNATIVES, ,
!t !
t o SCOPE i
j -
LMR'S AND HTGR '
}
i l
1
O O O PROPOSED PLANT CONFIGURATIONS-(RATINGS ARE APPR0XIMATE)
SAFR PRISM HTGR PLANT ELECTRICAL OUTPUT (MWE) 1,400 1,250 560 TG UNITS PER PLANT 4 3 2 l
REACTORS PER PLANT 4 9 4
- REACTORS PER TG UNIT 1 3 2 CONTROL ROOMS PER PLANT 1 1 1 l
1 i
REACTOR THERMAL RATING (MWT) 925 430 350 l
. i TG ELECTRICAL RATING (MWE) 350 420 280 EQUIVALENT ELECTRICAL RATING 350 140 140 i PER REACTOR (MWE) i l
~
O o o .
1 UTILITY PROBLEMS / ISSUES i- o PUBLIC ACCEPTANCE 4
o LICENSING PROCESS AND UNCERTAINTY o SHORTER AND MORE PREDICTABLE CONSTRUCTION TIME i
) o COST AND FINANCIAL RISK UNCERTAINTIES t
o IMPROVED PLANT CAPACITY FACTORS i .
,' o FINANCING / INSTITUTIONAL ALTERNATIVES
! o SilARED 0WNERSHIP RISKS i
]
i .
1 ,
i
p n v
a ks v
ADVANCED REACTOR CHARACTERISTICS o SMALL/ MODULAR O PASSIVELY SAFE o CERTIFIED AND STANDARDIZED PLANT O SHOP FABRICATION o PRE-ASSEMBLED - BARGE OR RAIL TRANSPORTABLE o COST COMPETITIVE o SHORT CONSTRUCTION TIME o HIGH PLANT CAPACITY FACTORS i
~
O o o NRC ADVANCED REACTOR POLICY STATEMENT REQUIREMENTS o RELIABLE, SIMPLE HEAT REMOVAL o EASY MAINTAINABILITY SYSTEMS o FEWER SUPPLEMENTAL SAFETY o REDUCED PERSONNEL EXPOSURE .
FEATURES o LONGER TIME CONSTANTS o MULTIPLE BARRIERS ,
o SIMPLIFIED SAFETY SYSTEMS o INCREASED STANDARDIZATION AND FABRICATION o UTILIZE INHERENCY, RELIABILITY, o EXPERIMENTALLY VERIFIABLE REDUNDANCY, DIVERSITY, SAFETY FEATURES 1 INDEPENDENCE o RELIABLE B0P EQUIPMENT i
.- o o ,9 ADVANCED REACTORS ACCOMPLISHMENTS TO DATE o LMR's/HTGR - MIDWAY THROUGH CONCEPTUAL DESIGN R
o NRC INTERACTION SCHEDULES ESTABLISHED AND UNDERWAY - POSITIVE EXPERIENCE o COSTS - PRELIMINARY ESTIMATES COMPETITIVE WITH. LWR's AND C0AL o SAFETY - ENHANCED PASSIVE CORE SHUTDOWN AND HEAT REMOVAL PROCESSES INCLUDED o R&D NEEDS EVOLVE FROM CONCEPTS - FOCUSES THE-SUPPORTING R8D PROGRAM o UTILITY INTERACTION - ACTIVELY PURSUED - FOSITIVE TO DATE
O
~
O O r 3 ACRS BRIEFING ON MHTGR AGENDA e PROGRAM STATUS e DESIGN, SAFETY, AND UCENSING APPROACH e DESIGN OVERVIEW e SAFETY CHARACTERISTICS L J VAXD:[WAUZEYJ10 13 7-WAR-86
r 7 l
l l MHTGR PROGRAM STATUS l DESIGN, SAFETY, AND LICENSING APPROACH PRESENTED TO THE ACRS MARCH 14,1986 A. C. MILLUNZI
. DEPARTMENT OF ENERGY I -
}
c J VAXD:[WAUZEY)10 7 11-WAR-86 l
1 - _ _ _ _ _ . _ _ _ _ _-
O O
O F 7 i
1 i
i
)
l PROGRAM OBJECTIVE L
i DEVELOP HTGR'S FOR BROAD RANGE OF
- APPLICATIONS IN SUPPORT OF COMMERCIAL / USER INTERESTS IN SAFETY AND HIGHER TEMPERATURE -
! CHARACTERISTICS OF THESE PLANTS. '
l l
l t
l
! L J VAXD:(WAUZEY]13 1 11-WAR-86
NRC INTERACTIONS I I FY 1985 l FY 1986 i FY 1987 LICENSING -
l PLAN PROCEDURAL q .
APPROACH '
l
, I l AGREEHENT -
AGRE MENT ON ON TO 2-LEVEL LICEN ING BASES
)
TECHNICAL
- APPROACH '
. I I l l 1
l l PSID i DESIGN / TECHNOLOGY I
-p
\7
! FAMILIARIZATION '
I I 1
l l LICENSABILITY l
l STATEMENT PSER l ,7
) DESIGN / TECHNOLOGY . <
l REVIEW
- I I i
l 1
1
l 0 o -
o l r 3 i
i i
PROCEDURAL APPROACH INTERACTION LICENS!NG PLANT DRAFT COMPLETE l NRC ADVANCED REACTOR l POLICY ISSUED FOR CONMENT . COMPLETE l
l LlCENSING PLAN FORMAL SUBMITTAL COMPLETE j INDUSTRY CONNENTS ON POLICY COMPLETE l
l NRC ACCEPTANCE OF LICENSING PL'AN COMPLETE l
c J l VA l13 2 11-WAR-86
O O O f 3 i
TECHNICAL APPROACH INTERACTION
! BRIEFING SUBMITTAL j TOP-LEVEL CRITERIA COMPLETE COMPLETE
! BRIDGING METHODS COWLETE 2/86 i
ACCIDENT SELECTION METHOD COWLETE 2/86 SAFETY CLASS SELECTION COMPLETE 2/86 PRINCIPAL DESIGN CRITERIA COWLETE 2/86 i
ACRS BRIEFING 1/86 NA l L J
! VA 113 3 (FUR:BE
o . -
o - -
o i 7 l DESIGN / TECHNOLOGY FAMlUARIZATION i
l lSSUE BRIEFING l MODULAR HTGR DESIGN 12/85
) FUEL 3/86 j DECAY HEAT REMOVAL 5/86 l REACTIVITY CONTROL 5/86 j CORE SUPPORT STRUCTURE 5/86 j ISI 7/86 i
WATER / AIR INGRESS 7/86 CONTAIbNENT/ CONF INEMENT 8/86 i BOP CLASSIFICATION 8/86 MULTIPLE MODULE CONTROL 8/86 i
STANDARD PLANT ISSUES 8/86 ACRS BRAEFING 9/86 c
YA l13 4 11-WAR-86
9 9 . .
9 r 7 f
i
! DESIGN / TECHNOLOGY FAMILIARIZATION (cont) l DOCUMENT DATE l TECHNOLOGY PLAN 9/86 I
PRA 9/86 PSID 9/86 l - OUTLINE COMPLETE l - FULL SUBMITTAL 9/86 i DESIGN AND TECHNOLOGY REVIEW ~
l PSER 6/87
! LICENSABiLiTY STATEMENT 9/87 l
N A VA H3 5 u-uAx-se
o 4
O
~ -
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HTGR DESIGN & LICENSING APPROACH i
I
\
USER TOP-LEVEL REGULATORY f
REQUIREMENTS I CRITERIA -
I : LICENSING BASIS
, r h
] ! PRINCIPAL DESIGN CRITERIA l INTEGR ATED APPRO ACH
, < BRIDGE N?
LICENSING BASIS EVENTS EQUIPMENT CLASS OTHER BASES ENGINEERING PRODUCT PLANT DESIGN, ETC. -
O O i O
- i 7 l
i i
i MODULAR HTGR I MEETS UTILITY USER REQUIREMENTS l e NOMINAL PLANT SIZE 500 MW(e) i i e EQUIVALENT AVAILABILITY >80%
i
-5 e PROBABlUTY OF PLANT LOSS <10 / MODULE-YR e MEET EXISTING SAFETY AND UCENSING CRITERIA WITH NO
-7 PUBUC SHELTERING FOR EVENT FREQUENCIES >5x10 /YR e 10% POWER COST ADVANTAGE OVER COAL i L J moSiuzmio ,4 7-u -.. .
O O
' ~
O F 7 l
i i
i i PROPOSED BASIS FOR l TOP-LEVEL CRITERIA SELECTION l 1. CRITERIA MUST BE DIRECT STATEMENTS OF ACCEPTABLE CONSEQUENCES OR RISKS i TO THE PUBUC OR THE ENVIRONMENT i 2. CRITERIA MUST BE INDEPENDENT OF PLANT DESIGN
- 3. CRITERIA MUST BE QUANTIFIABLE 1
L _
J VA 11 3 6 11- W AR-65
l 0
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l
! PROPOSED SOURCES AND CANDIDATES l FOR TOP-LEVEL REGULATORY CRITERIA
)
- NUREG-0880 l -INDIVIDUAL AND SOCIETAL MORTAUTY RISKS e 10CFR50 APPENDIX l 1 - NUMERICAL DOSE GUIDEUNES e 10CFR100
- - NUMERICAL DOSE GUIDEUNES o EPA-520
- PAG DOSES N Y YAXD:[WAUZEY]10 15 7-WAR-86
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PURPOSE OF THE INTEGRATED APPROACH l THE INTEGRATED APPROACH IS USED TO:
i i
o DEVELOP REQUIREMENTS o EVALUATE DESIGNS SELECTED TO MEET REQUIREMENTS l
l o COMMUNICATE l
i
? N A VAXD:[WAUZEYJ10 7 6-WAR-86 1
O O
' ~
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- F 7 i
l SAVINGS FROM THE INTEGRATED APPROACH i
l THE SAVINGS ENVISIONED FROM THE USE OF THE INTEGRATED j APPROACH ARE DUE TO:
l 0 A CLEAR UNDERSTANDING BY THE DESIGNERS, j CONTRACTORS, AND OPERATORS OF WHAT THEIR ROLES -
l AND RESPONSIBlUTIES ARE.
- e AN EARLY IDENTIFICATION OF INTERFACES WHICH l
REDUCES THE RISK OF LATER MORE COSTLY REVISIONS.
- 0 VISIBILITY OF THE BASIS FOR DESIGN REQUIREMENTS.
l 0 EUMINATION OF UNJUSTIFIABLE RETROFITS.
l 0 JUSTIFICATION FOR, OR DELETION OF, DEVELOPMENT PROGRAMS.
l L J 13 7 11-WAR-86
i o - -
o - -
o F .
3 i
i SAFETY PHILOSOPHY
! e PROVIDE DEFENSE-IN-DEPTH THROUGH PURSUIT OF l FOUR GOALS:
{ 1 - MAINTAIN SAFE PLANT OPERATION l 2 - MAINTAIN PLANT PROTECTION l 3 - MAINTAIN CONTROL OF RADIONUCUDE RELEASE l 4 - MAINTAIN EMERGENCY PREPAREDNESS
) e GOAL 1 TO BE ACHIEVED BY HIGHLY REUABLE j OPERATION AND WITH WELL TRAINED PERSONNEL e GOALS 2 AND 3 TO BE ACHIEVED THROUGH UTILIZATION OF INHERENT CHARACTERISTICS AND PASSIVE SAFETY FEATURES e GOALS 1 - 3 TO BE ACHIEVED SO \NELL THAT MINIMAL RELIANCE NEED TO BE PLACED ON GOAL 4 L J
= n3 . ii-vo-u
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i LICENSING METHODOLOGY
SUMMARY
e USE PRA TO IDENTIFY UKEUHOOD OF EVENTS AND CLASSIFY EVENTS INTO THREE REGIONS FOR COMPARISON AGAINST TOP LEVEL REGULATORY CRITERIA e EXAMINE EVENTS TO IDENTIFY REQUIRED FUNCTIONS TO MEET THE TOP LEVEL REGULATORY CRITERIA i i l- e CHOOSE DESIGN SELECTIONS TO ACCOMPUSH REQUIRED FUNCTIONS TO MEET THE TOP LEVEL REGULATORY CRITERIA e >
AP186.VWG 2 21-JAN-86
O O
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1 l
MHTGR SAFETY AND LICENSING APPROACH CONSISTENT WITH
. ADVANCED REACTOR POLICY e EARLY INTERACTION WITH NRC e CRITERIA SPECIFIC TO ADVANCED REACTORS e EMPHASIS ON INHERENT, PASSIVE SAFETY i
c >
VAXD:[MAUZEY)10 16 7-WAR-86 i ___ _ . . - . - _ _ - - -.
O O O
i i
i F 7 MODULAR HTGR DESIGN OVERVIEW l PRESENIED TO THE ACRS MARCH 14,1986 A. J. NEYLAN, DIVISION DIRECTOR 1
GA TECHNOLOGIES INC.
-,omeo.. , ,, _ ..
j 3 - - - - - - - - - - _
1 o - -
o - -
o i MODULAR HTGR
! KEY CHARACTERISTICS j
- BASIC HTGR FEATURES
- HEllUM GAS COOLANT -INERTlSINGLE PHASE l - GRAPHITE MODERATOR - LONG RESPONSE TIMES
- GRAPHITE CORE STRUCTURE - HIGH TEMPERATURE STABILITY l - CERAMIC COATED FUEL EMBEDDED IN GRAPHITE MATRIX - LOW RELEASESlLOW WORKER EXPOSURE
- SPECIAL MODULAR FEATURES
- CONFIGURATION SELECTED USING PASSIVE FEATURES ASSURES PUBLIC SAFETY
- STANDARDIZED, PRE-LICENSED, FACTORY-FABRICATED ASSEMBLY ,
O O
0 i
h 4-UNIT MODULAR HTGR PLANT i
i PLANT C' ONTROL ADMINISTRATION TURBlWE GENERATOR BUILDING BUILDING BUILDl4G FUEL HANDLING BUILDING g @
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REFUEllNG PENETRATIONS l
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, VESSEL l MAIN CIRCULATOR ACT R CORE '
350 MW(t)
MODULAR HTGR 31E.
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THE MODULAR HTGR HAS HIGH THERMAL EFFICIENCY I
NUCLEAR SYSTEM PARAMETERS THERMAL POWER, [4 x 350 MW(t)] 1400 HEllUM PRESSURE, PSIA 925 HEllUM TEMPERATURE,8F 497/1268 I
POWER SYSTEM PARAMETERS POWER CYCLE NON-REHEAT l TURBINE INLET PRESSURE / TEMPERATURE, PSIA /oF 2415/1000 -
SYSTEM PARAMETERS l NET ELECTRICAL OUTPUT, MW(e) 558 NET THERMAL EFFICIENCY, % 39.9 l
M#
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- LAMINAR" *BISO" *TRISO"
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O l.
TRISO-COATED HTGR l FISSILE AND FERTILE FUEL PARTICLES i OUTER ISOTROPIC l PYROLITIC CARBON ,
1 1
l SILICON CARBIDE i .
INNER ISOTROPIC
!; PYROLITIC CARBON
) - POROUS BUFFER 1 ,/
!}l
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. FISSILE PARTICLE FERTILE PARTICLE l
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h amm a a_ . . _ _ -
EVOLUTION OF HTGR FUEL QUALITY ,
1 10-3 *
- py NOTE: MEAN l -- UPPER l _ g 95% C.L
- 2240 e i
MW(t) TNTR MEAN T- " 50% C.L 1
i .
lii o
) (- 10-4 ,- e H08EG zE AVR GA (COMM)
- 6 X 10-5 i o lE -e- -
-- HOBEG --- 1984---o- - -1 r - --
! O l'E HR8 1981 DEV. MHTGR l HTR-100
- SPEC i
& "E HTR-500 "
) U+ INTERATOM H08EG
$b i 200 MW(e) 1984
, ag e
- I 5 10-5
, M _
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l I -
i 1 HOBEG l _ 1982 DEV.
]
l 10-6
! CASES H-552(1) 3-7-86
THE MODULAR HTGR ASSURES PUBLIC SAFETY i
l l.
- WITHOUT OPERATOR ACTIONS
- WITHOUT OPERATION OF ANY POWERED SAFETY SYSTEMS 1
- WITHOUT THE NEED OF PUBLIC SHELTERING OR EVACUATION
%'lA PASSIVE SAFETY
O
~~ '
O O ,
r 3 MHTGR SAi-t-1Y CHARACTERISTICS l
l l PRESENTED TO THE ACRS MARCH 14,1986 F. A. SILADY, SAFETY AND RELIABILITY MANAGER GA TECHNOLOGIES INC.
4 -
k )
VAXD:(WAUZEY]10 1 11-WAR-86 l '
O O
O 1 < ,
)
MHTGR SAtt: 1Y DESIGN OBJECTIVES f 1. PROVIDE FUEL QUAUTY AND SPECIFY NORMAL OPERATING CONDITIONS TO UMIT RADIONUCUDE INVENTORIES OUTSIDE OF FUEL TO MEET 10CFR100 DOSES AT PLANT l
I BOUNDARY
- 2. PROVIDE REUABLE PASSIVE DESIGN SELECTIONS
! TO RETAIN RADIONUCLIDES WITHIN FUEL TO MEET 10CFR100 DOSES AT PLANT BOUNDARY
- 3. PROVIDE ADDITIONAL DESIGN SELECTIONS TO RETAIN RADIONUCUDES WITHIN THE PLANT TO MEET PAG DOSES AT PLANT BOUNDARY l
l l k J VAXD:[WAUZEY]10 9 11-MAR-86
Q
\
_m g .._
MHTGR MEETS 10CFR100 BY RADIONUCLIDE RETENTION IN FUEL I
3.
gi; DENOTES FUNCTIONS NEEDED 3.;j TO MEET 10CFR100
^
j[Ni Ai$iS^d$$iN$[N!f
, l{ RAIN 0NUCLIOE RELEASEj!!
l
) 1 3.i r _ _ _ _t _ _ _ 3.2 i
[d6NiN6i" ! CONTROL PERSONNEL I
[ RAD;ATIONj! l ACCESS I l m_________J l 3.1.1 3.1.2 l 3.1.3
!!.!C6WVR6i'R4aiA idN!!! CONTROL RAMATION CONTROL RADIATION
{ FR0aA CORE FROM PROCESSES FROM STORAGE i I i l l 3.1.1.1 l 3.1.1.2 1 ' ...............:
CONTROL .
CONTROL ,.
DIRECT RADIATION 1RADIATl0N.TRANSPORTi!
. I i l 3.1.1.2.1 l 3.1.1.2.2 l 3.1.1.2.3 l 3.1.1.2.4
![U6NyN5['IRAMIE0Eili CONTROL TRANSPORT CONTROL TRANSPORT CONTROL TRANSPORT
- FROM CORE IN PRIMARY CIRCUIT FROM REACTOR BLDG FROM SITE
! I I
l 3.1.1.2.1.1 l 3.1.1.2.1.2
!iREYAiS'N INdNUddUi!N!! RETAIN RAMONUCUDES
. .I. N. . . .F.U E. .L. . P. AR. .T. .IC. LES.
. IN~CORE GRAPHITE H-558(1) 3-7-86 i
o - -
o - -
o i
f D COMPARISON OF l-131 ACTIVITY REDUCTION REQUIREMENTS .
- FOR THE MHTGR '
i
,l REOUiREhENT 10CFR100 (150 rsm) PAG (5 rem) i i
l A) ALLONABLE HUMAN UPTAKE 10-# Ci 3x10-8 Ci i
7 7 l B) CORE INVENTORY 10 Ci 10 Ci i
l C) PERMITTED UPTAKE 10- 3x10-'3
) FRACTION (A/B)
,1 l ATTENUATION FACTORS D) DISTANCE 4x10-7 4x10-7 (425 M EAB) (425 M EPZ)
E) RESIDUAL RELEASE 3x10-5 J1x10-8
. FRACTION PERMITTED .
l (C / D)
(i.e., considering fuel retention, vessel retention,
- plateout, settling,etc.)
\
i k ) .
] VAXD:[WAUZEYJ10 10 10-W AR-86 l
O O
~ '
! O i r m m aa - - __;m
! FUEL QUALITY LIMITS i PRIMARY CIRCUlT ACTIVITY i
RELEASE FUEL FRACTION LIMITED RELEASE FROM FUEL PARTICLES DURING NORMAL OPERATION '
- PLATE OUT ACTIVITY 1x10-8
- CIRCULATING ACTIVITY (10-8)
! SUBTOTAL 1x10-6
) (meets 10CFR100) l OTHER l
j LIFTOFF LIMITED BY LEAK AREA SIZE (4in Iine) .01 HELIUM BOUYANCY CAUSES ELEVATED RELEASE .03 TOTAL 3x10-8 (meets PAG) l VAXD:[WAUZEY]to 11 11-WAR-86 I
O O O REQUIRED FUNCTIONS TO RETAIN
! __ .._ RADIONUCLIDES IN FUEL PARTICLES RETAIN
! RADIONUCLIDES I IN FUEL PARTICLES i
i I
CONTROL REMOVE CONTROL CHEMICAL ATTACK CORE HEAT HEAT GENERATION 3$-8Y
O O O r 7 1
I
{ DESIGN SELECTIONS l TO REMOVE CORE HEAT i i
i e ADDITIONAL TO ACTIVE HEAT REMOVAL SYSTEMS (HTS and SCS) 1
- PASSIVE REACTOR CAVITY COOLING SYSTEM
- (RCCS) j - PRESSURIZED
- DEPRESSURIZED c J VAXD:{WAUZEY]10 2 8-WAR-86
O
~ '
O O
! r 7
\, _
l j PASSIVE CORE HEAT REMOVAL RESULTS i FROM MATERIAL CHARACTERISTICS AND DESIGN FEATURES j e SMALL THERMAL RATING
! - UMITS AMOUNT OF AF IERHEAT e CORE GEOMETRY
- - PROVIDES FOR REMOVAL OF AF IENHEAT BY PASSIVE CONDUCTION AND RADIATION l
- PASSIVE HEAT SINK
- UTIUZES REACTOR CAVITY C00UNG SYSTEM (RCCS) e SLOW HEAT UP OF MASSIVE GRAPHITE CORE e HIGH TEMPERATURE STABlUTY OF REACTOR CORE AND FUEL
- FISSION PRODUCTS RETAINED IN COATED PARTICLES L J v .uzmm ,2 .- - -..
O f ~ ~ '
O O
! z PRESSURIZED DECAY HEAT REMOVAL BY j o NATURAL CONVECTION AND RADIATION TO I - " ~ -
PASSIVE REACTOR CAVITY COOLING SYSTEM i .
l ,.J .l[ l "' l l""I I b
- 3 e:
, :;.:E l*
E ** TO EXHAUST p rgn
. ;;j FROM INTAKE
~
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. (4 LOCATIONS)
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jyj H-311(4)(A)
.j;s;!;..
2-19-86 (I-:. _ !.:?
O O
O i
DEPRESSURIZED DECAY HEAT REMOVAL l
4 BY CONDUCTION AND RADIATION TO 1
- " ~ "
PASSIVE REACTOR CAVITY COOLING SYSTEM s .=
$ff. I"l 1 "l -I.E i y:. i.t..g:..
3 j I.. d....-:.
TO EXHAUST l
FROM INTAKE W.'
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C0ATED FUEL PARTICLES MAINTAIN INTEGRITY TO VERY HIGH TEMPERATURES o
) 1.0 8-o-i TEST RESULTS FOR TRISO COATED UC0 0.8 - O FUEL PARTICLES i
E p O g 0.6 -
o i
E i w .
$ O.4 -
E
! NORMAL O.2 - PEAK FUEL OPERATING 0
l TEMPERATURE l
! O e I I I bg 3 l 1000 1200 1400 1600 1800 2000 2200 2400 2600 TEMPERATURE (8C)
N-150(9) 9-4-84
- ~
O O O
f 3 m.
PASSIVE RETENTION DURING
! TEMPERATURE TRANSIENTS l
1 RELEASE FUEL FRACTlON LIMITED HEATUP RELEASE FROM 2x10-8 FUEL PARTICLE COATINGS l
DELAYED RELEASE OF ACTIVITY (METEOROLOGY) .3 i
SUBTOTAL 6x10-6 (meets 10CFR100)
OTHER 1
I VESSEL HOLDUP AND RETENTION ONING .1 TO INHALATION AND DEPOSITION J
HOLDUP AND RETENTION WITHIN .01 REACTOR BUILDING TOTAL 6x10-8 (meets PAG)
VAXD:[WAUZEY]10 13 10-W AR-86 1
l _ _ . - - - _ - -
O
~
O O F 3
.l l
DESIGN SELECTIONS TO CONTROL CHEMICAL ATTACK BY WATER
) e UMITED SOURCES (MAGNETIC BEARINGS) e HIGH QUALITY FUEL (LOW HYDROLYSIS)
- REUABLE DETECTION AND ISOLATION (SINGLE LOOP) k )
VAXD:[WAUZEY]10 3 11-W AR-86
o -
o - -
o F 7 l _
j RETENTION DURING WATER INGRESS i
i RELEASE FUEL FRACTION NO HYDROLISIS OF FUEL PARTICLES milch
! DO NOT ALREADY HAVE FAILED COATINGS 2x10-#
i I
LIMITED RELEASE FROM PARTlCLES WITH FAILED COATINGS .015 SUBTOTAL 3x10-6 (meets 10CFR100) l OTHER l HYDROLISIS Or FUEL LIMITED BY AMOUNT
! OF WATER milch ENTERS CORE .2 l
RETENTION IN VESSEL AND AUXILIARY l BUILDINGS .1 i
TOTAL 6x10-8
] (meets PAG)
! k J VAXD:[WAUZEY]10 4 11-WAR-86 i'
~
O O O
~
1 h .-
LIMITED AIR-GRAPHITE REACTION RETAINS RADIONUCLIDES IN CORE 10-3 _
m :
m -
l 0 -
~
2
=
=
t: 104 :-
i 8E :
t 13 SQ. IN. LEAK i = at o8 -
/e hE -
/ 0.1 SQ. IN. LEAK I 58 10-5 i iii'il e i i i iiniil i i i iinii 1 2 jg3 10' 10 10 TIME (HRS) PAST j PRIMARY COOLANT LEAK LEADING TO CONDUCTION C00LDOWN i
H-558(11) 3-7-86 l
1 _________ __-- __. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
O O O r 3 PASSIVE RETENTION DURING AIR INGRESS l RELEASE FUEL FRACTION OXlDATION RELEASE LIMITED TO ACTIVITY 2x10-8 HELD UP IN GRAPHITE AFTER CORE HEATUP DELAYED RELEASE OF ACTIVITY (METEOROLOGY) .3 SUBTOTAL 6x10-6
! (meeis 10CFR100)
OTHER l
OXIDATION LIMITED BY PRIMARY COOLANT 1x10-4 BOUNDARY BREACH AND REACTOR BUILDING TOTAL 6x10-'
(meets PAG) l L J VAXD:[WAU'IEY]10 21 10-WAR-66
O O O F 3 Q __
l DESIGN SELECTIONS TO CONTROL HEAT GENERATION e FUEL NEGATIVE TEMPERATURE COEFFICIENT e CONTROL ROD SYSTEM e RESERVE SHUTDOWN SYSTEM S-VAXD:[WAUZEY]10 6 11-WAR-86
O O
O
_m g ._ _
ACCEPTABLE PRESSURIZED CONDUCTION C00LDOWN TEMPERATURES WITH AND WITHOUT REACTOR TRIP 1
1300 1200 -
MAXIMUM W/0 TRIP l [ MAXIMUM W/ TRlP m 1100 a
i IE h1000
-"[,_,,,_----------
AVERAGE W/0 TRIP j g '
$ 900 -/
i l
l
=
800 A f.l.------~~"~~~~~~'~"~''~~~*
i AVERAGE W/ TRIP .
i 700 j 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30
! TIME HRS I
- H-558(2) j 3-7-86 i
O O O
g ACCEPTABLE DEPRESSURIZED CONDUCTION C00LDOWN TEMPERATURES WITH AND WITHOUT REACTOR TRIP 1800
- 1600 -
MAXIMUM W/0 TRIP 1 o ~ ~ ~~~
i e MAXIMUM W/ TRIP ~~~~
j E 1400 -
y
" 1200 -
/,, ==~~ * * * ' " "
l h
- ,,/ . - .
j,, ,,,yS . ..N...=.-. . .. . . ._ = -:===:== ===
l h1000 AVERAGE W/0 TRIP l
800 h/'
r AVERAGE W/ TRIP
' ' ' ' ' ' ' ' ' ' ' ' ' I 600 O 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 i TIME HRS H-558(4) 3-7-86 i
F 7 i
DESIGN SELECTIONS TO CONTROL CHEMICAL ATTACK BY AIR i
e NON REACTING COOLANT j e HIGH QUALITY VESSEL e LIMITED SOURCE e EMBEDDED VESSEL l
l e EMBEDDED CERAMIC FUEL PARTICLES 1
i l
C J VAXD:[WAUZEY]10 5 11-WAR-86
r .
9 9 9 r 3
. . ~ . -
CONCLUSIONS e HIGH QUAUTY FUEL UMITS NORMAL OPERATION RADIONUCUDE INVENTORIES IN PRIMARY SYSTEM TO LEVELS WITHIN 10CFR100
i VAXD:[WAUZEY]10 14 11-MAR-86 a
O O O r 3
SUMMARY
e INHERENT CHARACTERISTICS AND PASSIVE FEATURES ASSURE I
RADIONUCLIDE RETENTION IN FUEL SUFFICIENT TO OBVIATE
- OFFSITE SHELTERING OR EVACUATION
]
- PROGRAM UTILIZES SYSTEMATIC, TRACEABLE UCENSING APPROACH SPECIFIC TO MODULAR HTGR CONSISTENT WITH ADVANCED REACTOR POUCY i
L J v4xo-[wAuztyJ10 22 11-WAR-86
j .
O O O
- l, I
4 I GDC-4 RULEMAKING l i t
5 i
SCHEDULE I
i 4
t 1
i PROPOSED FINAL 1
)
L I
LIMITED SCOPE RULE JULY 85 MARCH 86 BROAD SCOPE RULE MAY 86 DECEMBER 86 i
- i i .
l !
i t____________
> _ _ _ - - _ - __ ... .~ - - - _ .. - .- . . , . - _ .- .-
0' O o BROAD SCOPF GDC 14 Rul E RFSPONSE TO ACRS METALS SUBCOMMITTFE QUESTIONS
- 1. IS THE BROAD SCOPE Rule LIMITED TO HIGH ENERGY PIPING?
YES, THIS WAS THE INTENT. PIPE WHIP RESTRAINTS AND JET SHIELDS ARE ONLY LOCATED NEAR SUCH PIPING. THE SUPPLEMENTARY INFORMATION IN THE RULE IS MODIFIED TO STATE THAT ONLY HIGH ENERGY PIPING IS AFFECTED AND A DEFINITION OF HIGH ENERGY PIPING IS PROVIDED (SYSTEM PRESSURE GREATER THAN 275 PSIG OR OPERATING TEMPERATURE GREATER THAN 200 F)
- 2. ARE INDIVIDUAL BREAKS EXCLUDED OR ARE Ali BREAKS IN A FIUID SYSTFM FXClUDFD?
THE RULE SPEAKS OF THE " PROBABILITY OF FLUID SYSTEM PIPE RUPTURE" BEING EXTREMELY LOW. THE RULE DOES NOI DEAL WITH SPECIFIC BREAK LOCATIONS.
WHEN LEAK-BEFORE-BREAK IS APPLIED, IT IS APPLIED TO FLUID SYSTEM PIPING OR PORTIONS THEREOF. A PORTION OF A FLUID SYSTEM PIPING MEANS AN ANALYZABLE SEGMENT OF PIPING, USUALLY BETWEEN ANCHORS. THE RULE IS MODIFIED TO CLARIFY THIS.
~
O O o BROAD SCOPE GDC-4 RULE
- 3. WOUID ADDITIONAL ATTENTION BF GIVEN TO FLANGED JOINTS WHEN IFAK-RFFORF-BREAK IS APPLIED TO WELDED JOINTS?
NO. IN THE JUDGMENT OF THE STAFF, CATASTROPIC RUPTURES IN FLANGED JOINTS NEED NOT BE POSTULATED. PAST EVALUATIONS HAVE INDICATED THAT FLANGED JOINTS DO NOT RUPTURE. NO MODIFICATION NEEDED TO RESPOND TO THIS QUESTION.
- 4. WIll THE Rule REQUIRE AN EVALUATION OF CREEP RUPTURE IN PIPING?
YES. IN LWRS CREEP DOES NOT OCCUR AT THE TEMPERATURE AND FOR THE MATERIALS WHICH EXIST. HOWEVER, SINCE THE RULE APPLIES TO GAS AND METAL COOLED REACTORS, AN EVALUATION OF CREEP MAY BE REQUIRED IN SOME CASES. THE RULE IS MODIFIED TO REFLECT THIS COMMENT.
- 5. IS THE INF1UENCE OF PLUGGING INCLUDED IN THE LEAK RATE ESTIMATIONS?
N0, PLUGGING IS NOT CONSIDERED A PROBLEM FROM THE PRACTICAL POINT OF VIEW BECAUSE OF HIGH PRESSURES AND FLAW GE0METRIES (N0 TIGHT CRACKS);
HOWEVER, RESEARCH IS PLANNED AND ON-G0ING THAT ADDRESSES THIS ISSUE.
NO MODIFICATION NEEDED TO RESPOND TO THIS COMMENT.
1
BROAD SCOPE GDC-4 RULE
- 6. IF THERE ARE PUBLIC SAFETY BENEFITS. SHOULDN'T THE RULE BE MANDATORY RATHER THAN PERMISSIVE?
WHILE THE EFFECTIVENESS OF INSERVICE INSPECTION IS IMPROVED, AND PIPE RUPTURE PROBABILITIES ARE REDUCED (BECAUSE OF ELIMINATION OF INADVERTENT RESTRAINT OF THERMAL GROWTH) THESE BENEFITS HAVE NOT BEEN QUANTIFIED. AS A CONSEQUENCE, THERE IS NO SECURE BASIS FOR MAKING THE RULE MANDATORY. ON THE OTHER HAND, THE INCREASED PUBLIC RISKS ASSOCIATED WITH REMOVING DEVICES WHICH MITIGATE THE AFFECTS OF POSTULATED PIPE RUPTURES HAS BEEN SHOWN TO BE INSIGNIFICANT FOR REACTOR COOLANT LOOP PIPING.
O O O I
BROAD SCOPE GDC-4 RULE i
SAFETY ISSUE APPR0XIMATELY 15,000 PIPE WHIP RESTRAINTS IN SERVICE TODAY OF VARYING SIZES AND DESIGNS (INSTALLED AT A COST OF ROUGHLY $2 BILLION [1985 D0LLARS) IN DIRECT COSTS; IF STRETCHED OUT SCHEDULE AND FINANCE COSTS ARE INCLUDED, FIGURE JUMPS TO $6 BILLION), MORE THAN 100,000 EXPERIENCE YEARS ACCUMULATED WITH PIPE WHIP RESTRAINTS, WITHOUT HOWEVER, EVEN ONE INSTANCE WHERE A PIPE WHIP RESTRAINT WAS NEEDED TO INDIAN POINT, DUANE ARNOLD AND MAINE PERFORM A SAFETY FUNCTION.
YANKEE, WHICH TO DATE REPRESENT THE MOST SEVERE LEAK / BREAK PIPING PROBLEMS IN SERVICE, DID NOT REQUIRE PIPE WHIP RESTRAINTS TO MAINTAIN SAFETY. WE BELIEVE FOREIGN EXPERIENCE.IS SIMILAR.
PIPE WHIP RESTRAINTS CAN DEGRADE SAFETY BY LIMITING THERMAL EXPANSION WHEN CONTACT BETWEEN THE PIPE WHIP RESTRAINT AND PIPE INADVERTENTLY OCCURS DUE T0:
4 A. INTERACTION WITH ANY OF THE ESTIMATED 2000 LOCKED (FAILED)
SNUBBERS NOW INSTALLED (APPROXIMATELY 40,000 SNUBBERS IN SERVICE, WITH 4000 FAILED " FREE" AND 2000 FAILED " LOCKED").
s
O O O BROAD SCOPE GDC-4 RULE SAFETY ISSUE (CONTINUED)
B. INABILITY TO LIMIT, CONTROL OR ESTIMATE THE INFLUENCE ON THERMAL GROWTH OF PIPE SUPPORT GAPS.
C. DIFFICULTIES WITH MAINTAINING TOLERANCES AND ALIGNMENTS IN PIPE WHIP RESTRAINTS, WHICH IN THE COLD CONDITION MAY BE SEVERAL INCHES AWAY FROM PIPING, WHILE DURING OPERATION ARE TO BE A FRACTION OF AN INCH AWAY IN SOME CASES.
- THESE CONTACT PROBLEMS MAY CAUSE CRACKS TO GROW AT LOCATIONS NOT NOW PROTECTED AGAINST PIPE RUPTURE BECAUSE OF MODIFIED STRESSES.
I PIPE WHIP RESTRAINTS LIMIT ACCESSIBILITY FOR AND DIMINISil l EFFECTIVENESS OF INSERVICE INSPECTION WHILE INCREASING WORKER RADIATION EXPOSURES.
l l .- -.
O O o BROAD SCOPE GDC-4 RULE REASONS FOR LIMITING LBB TO DYNAMIC EFFECTS 0 LEAKS, VALVE MALFUNCTIONS AND OTHER SOURCES OF BLOWDOWN IMPOSE REQUIREMENTS FOR CONTAINMENTS, ECCS AND ENVIRONMENTAL QUALIFICATION WHICH CANNOT BE ELIMINATED BY INVESTIGATING AND DEMONSTRATING PIPING INTEGRITY.
O IF LBB IS APPLIED TO THE CONTAINMENT, ECCS AND ENVIRONMENTAL QUALIFICATION DESIGN BASES TO ELIMINATE THE DEGB ACCIDENT, THEN A REPLACEMENT PIPE RUPTURE ACCIDENT MUST BE DEVELOPED FOR THESE ASPECTS OF FACILITY DESIGN, NONE NOW EXISTS.
O MAJOR LONG TERM RULEMAKING WOULD BE NEEDED TO ADDRESS OTHER THAN DYNAMIC EFFECTS, THEREBY FORESTALLING IMMEDIATE PAYOFF IN ELIMINATING PIPE WHIP RESTRAINTS, JET IMPINGEMENT BARRIERS AND UNDERTAKING OTHER BENEFICIAL FACILITY MODIFICATIONS.
O DYNAMIC EFFECTS LEAD TO THE PLACEMENT OF COUNTER PRODUCTIVE HARDWARE WHICH NEGATIVELY AFFECTS PLANT PERFORMANCE IN TERMS OF DEGRADED SAFETY AND INCREASED COSTS. OTHER ASPECTS OF FACILITY DESIGN MAY BE NEGATIVELY AFFECTED BY POSTULATED DEGB'S, BUT NOT TO THE SAME DEGREE THAT DYNAMIC EFFECTS IMPOSE COUNTER PRODUCTIVE REQUIREMENTS.
.O O O 3
BROAD SCOPE GDC-4 RULE DEFINITION OF DYNAMIC EFFECTS 0 PIPE WHIP AND OTHER PIPE BREAK REACTION FORCES.
O JET IMPINGEMENT FORCES.
'O DECOMPRESSION WAVES WITHIN THE RUPTURED PIPE (AFFECTS PIPE COMPONENTS AND THEIR INTERNALS).
O PRESSURIZATION IN CAVITIES, SUBCOMPARTMENTS AND COMPARTMENTS, EXCEPT WHEN THESE VOLUMES ARE PART OF THE CONTAINMENT SYSTEM.
O PIPE RUPTURE GENERATED MISSILES (INSULATION, PIPE SUPPORT BOLTS, ETC.).
l
BROAD SCOPE GDC-4 RULE 1 1
(SOURCE:
INTERNATIONAL LBB SEMINAR ON OCT 28-30, 1985)
- FOREIGN PRACTICES:
GENERAL NOTES: 1. N0 NATION ACCEPTING OR CONSIDERING LBB AT THIS TIME WILL MODIFY REQUIREMENTS FOR CONTAINMENTS, ECCS OR ENVIRONMENTAL QUALIFICATION. !
J 1
', CRITERIA FOR LEAKAGE DETECTION.
UK STRONG INCLINATION TO REJECT LBB FOR SIZEWELL BASED ON 1 CONCERNS WITH SCC AND NDE (CEGB AT'0DDS WITH NII ON THIS).
FRANCE UNDECIDED, BUT WEAKLY INCLINED TO REJECT LBB AT THIS TIME l
PARTLY BECAUSE OF COMMITMENT TO STANDARDIZATION. RESEARC ;
,I LBB IN PROGRESS. i FRG STRONG COMMITMENT TO LBB IN PWR MAIN COOLANT, MAIN FEED AND MAIN STEAM INSIDE CONTAINMENT. t i
i ITALY CLOSE TO FRG PRACTICES.
i i
I i
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_ _ _ _ _ . . _ . . _ _. _ _ _ _ ._ _ . _ _ _ _ _ _ _. _ _ _ _ .._ _____.. _ __ _ . _ _ _ -. _ . _ ___ _ _ _ _ ._ __ .. ..~ _ . . . _ . _
O O O 1
l BROAD SCOPE GDC-li RULE (CONTINUED)
) FOREIGN PRACTICES:
JAPAN INCLINED TO ACCEPT LBB, HEAVY INVESTMENT IN LBB RESEARCH,
! CANADA INCLINED TO ACCEPT LBB FOR CERTAIN PIPING SYSTEMS AT THE DARLINGTON FACILITY.
MOST OTHER COUNTRIES SEEM TO BE LESS ACTIVELY INVOLVED IN LBB, OR ARE:
OBSERVING DEVELOPMENTS IN THE LEADERSHIP COUNTRIES AB0VE.
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Comparisen Between U.S. and FRG Proposed Revistrns far ploe Rupture Desian Reeutrements at nuclear Power rients O
Ecope of piping Limited to pWR main coolant loop, pressuriser Any piping in any reactor surge line, main stese lines, feedwater line, type which meets rigorous RNR lines and other high energy lines inside acceptance criteria, piping the containment. SWR piping not included. subject to fatiure from water piping outside containment nopncluded. hasser, corrosion, indirect sources of rupture and high and low cycle fatigue ex-cluded from consideration.
Only pWR primary coolant loops known to satisfy rigorous acceptance criter-ja at this time.
Old piping vs. Limited to new piping which has several Age not a factor, piping in new piping quality taproveeents, such as forged operating plants, plants fittings, double walled piping, lower under construction and stress levels, better in-service inspection, future designs included.
, snubbers replaced with bumpers on components and high standards for leak detection.
Requirements for Not affected. Not affected.
emergency cora cooling systems, environmental qualification and O containment pressurination Dynamic loads on Retained in modified fore; previous dynamic Eliminated.
heavy component requirements replaced with a static analysis
, supports from with a safety factor of 2.
pipe rupture Replacement for 101 flow area longitudinal or circumferential None; however, breaks may double-ended break or some other requirement based on still be postulated in pipe rupture for fracture mechanics analyses. connecting branch pipes.
(ynamic loads pressurization Retained. Eliminated except in volumes in compartments, related to the containment j subcompartments function.
and cavities from
) pipe rupture i
I Effects of Double-ended pipe twpture eliminated in pri- Eliminated.
decompression mary coolant loop; however, double-ended pipe waves on heavy rupture retained in main steam and feedwater component inter- lines for the steam generators. -
nals Leakage detec- Leakage detection depended upon, and described Must be reliable, redundant, tion in detail in " Appendix 1 of Draft of Trans- diverse and sensitive so that actions of 173rd Meeting of RSK." a margin greater than lo on
! detection of unidentified l 1eakage from throughwall flaws exists.
O O O BROAD SCOPE GDC-4 RULE VALUE/ IMPACTS 0 FOR THE LIMTIED SCOPE RULE, APPLYING LBB TO PRIMARY COOLANT LOOP PIPING IN A FORECASTED POPULATION OF 85 PWRS LED TO THE FOLLOWING BEST ESTIMATE RESULTS (BASED SOLELY ON PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS):
AVERTED RADIATION EXPOSURES: 34,000 MAN-REM REDUCED COSTS: $186 MILLION HOWEVER, HEAVY COMPONENT SUPPORT REDESIGN CAN YIELD EVEN GREATER VALUE/ IMPACTS. FLORIDA POWER CORP. IS ESTIMATING COST SAVINGS OF
$20 MILLION AND AVERTED RADIATION EXPOSURES OF 2000 MAN-REM DUE SOLELY TO REDESIGN OF REACTOR COOLANT PUMP SUPPORTS AT CR-3.
O IF LBB COULD BE APPLIED TO BWR RECIRCULATION LOOP PIPING UNDER THE BROAD SCOPE RULE, THE FIGURES AB0VE COULD BE INCREMENTED AS FOLLOWS (BASED ON A FORECASTED 38 BWRS AND CONSIDERING ONLY PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS):
AVERTED RADIATION EXPOSURES: 8,600 MAN-REM REDUCED COSTS: $30 MILLION 0 FOR ANY FUTURE PLANT, COST SAVINGS OF APPR0XIMATELY $100 MILLION ARE ESTIMATED PER PLANT THROUGH EXCLUSION OF PIPE BREAKS (DUE TO IMPROVED CONSTRUCTION SCHEDULES, REDUCTIONS IN FINANCING COSTS AND DIRECT DESIGN AND CONSTRUCTION COSTS). SOME OF THESE PIPE BREAKS CAN BE EXCLUDED VIA THE PROPOSED REVISION TO SRP 3.6.2.
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O O BROAD SCOPE GDC-4 RULE NRC RESOURCE DEMANDS 0 GE (FOR GESSAR) AND DUQUESNE LIGHT CO. (FOR BEAVER VALLEY, UNIT
- 2) ARE ALREADY MAKING DEMANDS FOR NRC RESOURCES UNDER THE BROAD SCOPE RULE.
O IT IS EXPECTED THAT IF THE BROAD SCOPE RULE IS NOT PUBLISHED THE COMMISSION WILL BE INUNDATED WITH REQUESTS FOR SYSTEM AND PLANT UNIQUE EXEMPTIONS TO GDC-4 IN ADDITION TO PETITIONS FOR
, RULEMAKING. THIS EXPECTATION ARISES BECAUSE THE PIPING REVIEW COMMITTEE PUBLICALLY SUPPORTED A BROAD SCOPE RULE AND THE LIMITIED SCOPE RULE STATED THAT "THE COMMISSION WILL PROPOSE A TWO-THIRDS OF COMMENTERS ON THE BROADER AMENDMENT TO GDC-4".
LIMITED SCOPE RULE URGED THAT THE BROADER RULE BE EXPEDITED.
i
! O APPR0XIMATELY TEN TO TWENTY NRC MAN YEARS OF EFFORT PLUS
! ADDITIONAL RESEARCH OVER THE NEXT SEVERAL YEARS ARE ESTIMATED TO RESPOND TO INDUSTRY INITIATIVES TAKEN UNDER THE BROAD SCOPE RULE.
l NOTE: INDUSTRY HAS ALREADY EXPENDED AND IS CONTINUING TO EXPEND
- CONSIDERABLE RESOURCES FOR LBB RESEARCH, ANALYTICAL TECHNIQUES AND CRITERIA DEVELOPMENT.
O O O BROAD SCOPE GDC-4 RULE RELATED NRC REGULATORY ACTIONS SRP 3.6.2 IS BEING REVISED TO ELIMINATE REQUIREMENTS FOR ARBITRARY INTERMEDIATE PIPE BREAKS. THIS REVISION WILL ALSO ALLOW THE REMOVAL OF PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS, BUT IS HOWEVER, THE SRP REVISION HAS ABOUT INDEPENDENT OF THIS RULEMAKING.
THE PIPE RUPTURES THE SAME POTENTIAL MAGNITUDE OF VALUE/ IMPACTS.
BEING ELIMINATED VIA THE SRP REVISION ARE:
- 1. ARBITRARY IN THE SENSE THAT THEY ARE REQUIRED WITHOUT PHENOMEN0 LOGICAL BASES, THAT IS, EVEN THOUGH STRESSES AND USAGE FACTORS (A FATIGUE MEASURE) ARE ACCEPTABLY LOW, THESE BREAKS ARE STILL POSTULATED,
- 2. INTERMEDIATE TO CONTRAST WITH TERMINAL END BREAKS WHICH ARE STILL POSTULATED. TERMINAL END BREAKS ARE POSTULATED WITHOUT SPECIFIED PHENOMEN0 LOGICAL BASES BECAUSE HISTORY TEACHES THAT PIPE RUPTURES ARE MORE LIKELY TO OCCUR AT TERMINAL ENDS.
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O O O BROAD SCOPE GDC-fi RULE RELATED NRC REGULATORY ACTIONS (CONTINUED)
THIS RULEMAKING AFFECTS POSTULATED BREAKS WHICH ARE NOT NORMALLY ARBITRARY INTERMEDIATE BREAKS. ABOUT FIFTEEN NUCLEAR POWER UNITS HAVE ALREADY BEEN ALLOWED TO RELAX ARBITRARY INTERMEDIATE BREAK REQUIREMENTS EVEN THOUGH THE REVISION IS STILL IN PROGRESS.
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"O THIS PACKAGE DOES NOT INCLUDE THE REGULAT0 TORY ANALYSIS l
For: The Comissioners From: Victor Stello, Jr.
Acting Executive Director for Operations .
t l
Subject:
PROPOSED BROAD SCOPE RULE TO MODIFY GENERAL DESIGN CRITERION 4 0F 10 CFR PART 50 l
Purpose:
To obtain Comission approval to publish a notice of proposed rulemaking.
Category: This paper covers a major policy issue. Resource estimates Category I, preliminary.
l' Issue: Research, probabilistic assessments, and field experience fety caa 6. a 9 tiv ir ff ct e 61 i
O inaic t- that hardware installed to resist the dynamic effects from certain postulated pipe ruptures. This action expands the scope of affected piping in a recent proposed modification to General Design Criterion 4 (GDC 4) which allowed i- exclusion of dynamic effects ' associated with postulated
' pipe ruptures water in only) primary reactors (PWRs. coolant As in the loops limited scope of rule, pressurized
- non-mechanistic pipe rupture is still postulated as the
{ design basis for emergency core cooling systems, '
i containments, and environmental qualification. Operating plants, plants under construction, and future plant designs i are affected by this action.
Sumary: A limited scope amendment to GDC 4 was proposed to the Comission in SECY-85-108, dated March 26, 1985. This amendment allowed use of analyses to exclude pipe ruptures in the primary coolant loop piping of PWRs. An extension of this limited application to those high energy piping
- systems in all nuclear power units that meet rigorous
- acceptance criteria is now being recomended in this I
Contact:
J. A. O'Brien, RES 443-7854 K. R. Wichman, NRR I
i 492-9430
! W. M. Shields, OELD i 492-8693
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{ The Commissioners lO i
paper. The amendment to GDC 4 now recomended would permit a potentially more extensive removal of protective devices i
such as pipe whip restraints and jet impingement shields originally designed to mitigate the dynamic effects of postulated instantaneous pipe ruptures and other related changes.
3
Background:
The two-step approach to rulemaking was adopted because safety and economic benefits could immediately be obtained by an amendment limited to the primary loops of PWRs.
Sufficient technical evidence had already been developed by the NRC and industry to support the limited application of leak-before-break technology to PWR primary loops. This information, along with operating experience, had been reviewed and accepted as an adequate basis for decision making by the NRC staff, the ACRS and the CRGR.
i l
Additionally, a number of applicants and licensees had i requested exemptions within the purview of the limited scope rule. While the limited rule was based on (but not limited to) the alternative resolution of USI A-2 l
O previously reviewed and endorsed by the ACRS and the CRGR, the broad scope rule herein recommended by the staff 1 required completion of ongoing research and additional j'
scrutiny by these two bodies, particularly with respect to the acceptance criteria which would be applied to piping other than the primary loops of PWRs. These acceptance j
criteria for applying leak-before-break analyses are enunciated by the NRC staff in NUREG-1061 Volume 3, November 1984. The acceptance criteria were approved by l
the USNRC Piping Review Committee, but were not formally j reviewed at that time by the ACRS and the CRGR.
Discussion: General Design Criterion 4 as applied in the context of the i
definition of " loss-of-coolant" accident has required the
! installation of protective devices in nuclear power plants to mitigate events which are now regarded as extremely unlikely in piping which meet rigorous acceptance criteria.
This conclusion is based on the results of research, insights from probabilistic analyses and licensing and field experience. These protective devices impede inservice inspection and maintenance, reduce safety if improperly installed, and increase worker radiation
' exposures. The net contribution to safety from these devices is judged to be negative. To overcome these l
difficulties, the addition of two sentences to GDC 4 is l
proposed which allows the use of analyses to demonstrate piping integrity in all high energy piping in all nuclear power units. As a minimum, a fracture mechanics evaluation
9 The Comissioners 3 O
including the effects of fatigue is undertaken. Evaluation of potential water hammer, corrosion, creep and indirect failure mechanisms which could lead to pipe rupture are also required.
For existing PWRs, considering primary coolant loops only, cost savings of $186 million and reductions in worker exposures of 34,000 man-rem are estimated for a population of 85 PWRs. These figures did not include savings resulting from redesign of heavy component supports. One a
licensee seeking to take advantage of the modification of GDC 4 is estimating a per plant cost savings of $20 million and reduced worker exposures of about 2000 man-rem associated solely with a redesign of reactor coolant pump supports.
The above-mentioned value-impacts were realized under the !
]
4 already published limited scope amendment to GDC 4.
I Additional benefits which can be achieved under this broader amendment are discussed below.
For existing BWRs. considering only the recirculation loop O
piping, cost savings of $30 million and reductions of 8,600 worker man-rem are estimated for a population of 38 plants.
In existing PWRs and BWRs, public risk is estimated to be insignificantly impacted, or if credit is taken for i improved inservice inspection and enhanced safety, public risk is reduced by an unquantified amount.
l In existing plants, the staff has not quantified situations other than those discussed above; however, it is believed that other high energy piping will also indicate favorable value-impacts.
Value-impacts resulting from this rule are greatest for future plants, where estimated costs can be reduced approximately $100 million per unit. Of this sum, about
$30 million are direct costs and the balance stems from reduced financing costs and improved scheduling. Reduction in worker radiation exposures vary from plant to plant, but are in the range of 300 to 800 man-rem. Public risk was not quantified, but is believed to decrease due to improved effectiveness of inservice inspection and enhanced safety.
The above quoted figures are based primarily on the ,
j elimination of pipe whip restraints and jet imp.ingement ,
barriers and do not treat other facility changes that could 1 result from this rule.
1 This rulemaking will introduce an inconsistency into the l design basis for certain regulations by excluding only the 1
1
l The Comissioners 4 O
dynamic effects of postulated pipe ruptures while still retaining non-mechanistic pipe rupture as the design basis for emergency core cooling systems, containments and environmental qualification. The staff recognizes the need to address whether and to what extent leak-before-break analysis techniques may be used to modify present requirements relating to other features of facility design.
However, this is a longer term evaluation. For the present, the proposed rule allows the removal of plant hardware which it is believed negatively affects plant performance, while not affecting emergency core cooling systems, containments, and environmental qualification.
The staff plans to begin development of a Regulatory Guide and Standard Review Plan section dealing with the proposed acceptance criteria associated with the Broad Scope Rule to modify GDC 4 after review and evaluation of solicited public comment have been completed.
Recomendation: That the Comission:
- 1. Approve publication of a notice of the proposed O- amendments in the Federal Register (Enclosure 1);
- 2. Certify that the proposed rule, if promulgated, will not have a significant economic impact on a substan-tial number of small entities. This certification is necessary in order to satisfy the requirements of the Regulatory Flexibility Act, 5 U.S.C. 605(b); and
- 3. Note:
- a. The notice of proposed rulemaking in Enclosure I will be published in the Federal Register allowing 60 days for public coment.
- b. The Chief Counsel for Advocacy of the Small Business Administration will be informed of the certifications and the reasons for it as required by the Regulatory Flexibility Act.
- c. The proposed rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Managemer.t and Budget approval number 3150-0011.
The Consnissioners 5 O
- d. The Federal Register notice of proposed rule-making will be distributed to affected licensees and nonlicensees,
- e. A regulatory analysis (Enclosure 2) has been prepared for this rulemaking.
- g. A backfit analysis (Enclosure 3) has been prepared in accordance with 10 CFR 50.109.
Victor Stello, Jr.
Acting Executive Director for Operations O
Enclosures:
- 1. Federal Register Notice i 2. Regulatory Analysis
- 3. Backfit Analysis
- 4. Public Announcement
- 5. Congressional Letters
- 6. Environmental Assessment I
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NUCLEAR REGULATORY COMMISSION
'nv 10 CFR Part 50 Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures AGENCY: Nuclear Regulatory Comission.
ACTION: Proposed rule.
SUMMARY
- The Nuclear Regulatory Commission is proposing to expand the scope of a previous amendment to its regulations dealing with the O Protectioa e structures, syst-s and c- nents im rtant to safety against dynamic effects from postulated pipe ruptures. A recent proposed rule (50 FR 27006, July 1,1985) was limited to the primary coolant loop piping of pressurized water reactors (PWRs), whereas this present action would cover all high energy piping in all nuclear power plants. This expanded modification of General Design Criterion 4 (GDC 4) is based on the results of recent research and insights from probabilistic risk analyses, and would allow demonstration of piping integrity by analyses to serve as a basis for excluding consideration of dynamic effects associated with pipe ruptures. A more extensive removal of counter productive hardware, such as ;
i pipe whip restraints and jet impingement shields, and other related design changes would be permitted if rigorous acceptance criteria are satisfied.
The amendment will not impact other design requirements such as emergency core cooling system (ECCS) performance, containment design, and environmental qualification.
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[7590-01]
O o^1E: Comeat periee expires c60 dass efter pubiscation). Coments received after this date will be considered if it is practical to do so, but assurance of consideration can only be given to comments received on or before this date.
ADDRESSES: Send coments to: The Secretary of the Comission, U.S.
Nuclear Regulatory Comission, Washington, DC 20555, ATTN: Docketing and Service Branch.
Deliver coments to: Room 1121, 1717 H Street, NW, Washington, DC between 8:15am and 5:00pm weekdays.
Copies of the regulatory analysis, documents referenced in this notice, and coments received may be examined at: the NRC Public Document Room at 1717 H Street, NW, Washington, DC.
FOR FURTHER INFORMATION CONTACT: John A. O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone (301)443-7854.
SUPPLEMENTARY INFORMATION:
Table of Contents
, I. Background II. Scope of Rulemaking III. Proposed Rule IV. Sumary of Acceptance Criteria V. Invitation to Coment O vi Avaiiabiiity or oocume#ts VII. Finding of No Significant Environmental Impact: Availability VIII. Paperwork Reduction Act Statement 2
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[7590-01]
- IX. Regulatory Analysis O X. Backfit Rule XI. Regulatory Flexiblity Act Certification
' XII. List of Subjects In 10 CFR Part 50 BACKGROUND Background to this rulet king can be found in the limited scope modification to GDC 4 published as a proposed rule in the Federal Register on July 1, 1985 (50 FR 27006). Research performed by the NRC and industry, coupled with operating experience, have indicated that safety can 4 be negatively impacted by the placement of protective devices such as pipe whip restraints near certain piping. The Comission adopted a two-step approach to the modification because safety and economic benefits could be Q
quickly realized without extensive and time consuming review and discussion if the scope were initially limited to the primary main loop piping of PWRs. Substantial evidence had already been developed to show that the
" leak-before-break" concept was va'id for primary main coolant loops of PWRs. The Comi.;s',on decided not ,o defer the limited application of leak-before-break technology while the detailed provisions of the proposed acceptance criteria were being reviewed and approved. Many near term operating license (NT0L) nuclear power plant units and operating nuclear power plant units had requested exemptions from the requirements of GDC 4 and could benefit from the limited scope rule. A broader application of leak-before-break technology requires adoption of the proposed general ,
acceptance cr'iteria published in NUREG-1061', Volume 3, Chapter 5, November O 1984, entitled " Report of the U.S. Nuclear Regulatory Comission Piping Review Comittee, Evaluation of Potential for Pipe Breaks".
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[7590-01]
SCOPE OF RULEMAKING i -O
! The direct dynamic effects of pipe rupture are missile generation, pipe i whipping, pipe break reaction forces, jet impingement forces, decompression waves within the ruptured pipe and prassurization in cavities, subcompartments and compartments.
i To retain high safety margins, the application of leak-before-break technology to various piping systems should not decrease the capability of containments to perform their function of 3solating the outside environment from potential leaks, breaks, or malfunctions within the containment.
Containments will continue to be designed to accommodate loss of coolant accidents resulting from breaks in the reactor coolant pressure boundary up to and including a break equivalent in si.ze to the double-ended rupture of Q ,
the largest pipe. in the reactor. coolant system. As a consequence, pressurization from such breaks of the volumes indicated below are still included in the design basis even though other pressurizations due to pipe
. ruptures are excluded:
- 1. The containment pressure boundary,
! 2. Channeling elements (vents, vent headers, and downcomers) in BWR pressure suppre,sion type containments,
- 3. Compartdents necessary to the containment function, for example, PWR ice condenser containment structures providing separation of containmen't volumes.
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.,y. - ,- - _ . , . - . . . - . .
--. y 3,~--,%...r. ..4. -.._,..my--,-e_-..-r-,,-r.- w
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[7590-01)
Likewise, the design bases for emergency core cooling systems, and for environmental qualification still retain non-mechanistic pipe rupture.
However, further investigations may indicate a need to modify these requirements as well.
This proposed amendment to GDC 4 allows exclusion from the design basis of dynamic effects associated with high energy pipe rupture by application of
,4 leak-before-break technology. Only high energy piping in ruclear power units that meet rigorous acceptance. criteria are covered. High energy piping is defined as those systems having pressures exceeding 275 psig or temperatures exceeding 200 F.
Studies completed by Lawrence Livermore National Laboratory under contract O to the NRC indicate that adverse safety implications can result from requiring protective devices to resist the dynamic effects associated with postulated pipe rupture. (See NUREG/CR-4263, " Reliability Analysis of Stiff Versus Flexible Piping, Final Project Report", May 1985). The placement of pipe whip restraints degrades plant safety when thermal growth is inadvertently restricted, reduces the accessibility for and effectiveness of inservice inspections, increases inservice inspection i radiation dosages and adversely affects construction and maintenance economics. l When leak-before-break is shown to be applicable to a fluid system piping, pipe-connected component supports may be redesigned excluding the effects O of nine rupture. The use of 45aE code aiiowabies in component suonort redesign is judged sufficient for preventing pipe rupture due to component 5
[7590-01]
support failure. Redesigned component supports must have sufficient margins such that component support failure is a remote cause of pipe rupture.
Also, leak-before-break technology is only applicable to an entire piping system or analyzable portion thereof. Analyzable portions are typically segments located between anchor points. Leak-before-break cannot be applied to individual weld joints or other discrete locations.
PROPOSED RULE The proposed rule consists of two substitute sentences at the end of GDC 4 (replacing the sentence introduced by the limited scope rule) permitting the use of analyses to exclude dynamic effects of pipe ruptures in all high energy piping in all nuclear power units. As a minimum, a deterministic O fracture mechanics evaiuation inciuding the effects of fatigue is undertaken. Evaluations of potential water hamer, corrosion, creep and indirect failure mechanisms which could lead to pipe rupture are also required. In order to demonstrate that the probability of fluid system piping rupture is extremely low, applicants and licensees may follow procedures and acceptance criteria developed by the staff. Current procedures and acceptance criteria are presented in Chapter 5 of Volume 3, NUREG-1061, November 1984. These acceptance criteria are sumarized in the next section.
The supporting safety analysis must demonstrate from the results of a fracture mechanics analysis that a substantial range of stable pipe crack sizes can exist for an extended period which provide detectable leaks, and that the fluid systems piping will not rupture under these conditions consistent with the design basis for the piping.
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[7590-01]
The language of the proposed rule specifies " conditions consistent with the design basis for the piping." The design basis for the piping means those conditions specified in the FSAR, as amended, and may include 10 CFR Part 50 (especially the General Design Criteria in Appendix A to Part 50),
applicable sections of the Standard Review Plan, Regulatory Guides and industry standards such as the ASME Boiler and Pressure Vessel Code.
The heading of GDC 4 is revised from " Environmental and missile design bases" to read " Environmental and dynamic effects design bases" to clarify that this General Design Criterion covers other dynamic events than the effects of missiles.
The tenn " extremely low" is used in this amendment to GDC 4 with reference
. to the probability of fluid system pipe rupture. For reactor coolant loop piping, a representative value which would qualify as " extremely low" would be of the order of 10-0 per reactor year when all rupture locations are considered in the fluid system piping or portions thereof. For other piping, representative values will be developed consistent with this definition as the need arises. Alternatively, a deterministic evaluation with verified design and fabrication, in addition to adequate inservice l inspection, can meet the extremely low probability criterion. The deterministic evaluation is based on the requirement that structures and components are correctly engineered to meet the applicable regulations and NRC-endorsed industry codes.
Modifications of the licensed plant design of operating plants may involve l
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an unreviewed safety question under 10 CFR 50.59. Licensees of operating f plants desiring to make modifications may submit a license amendment for NRC approval in accordance with revised General Design Criterion 4. The license amendment may also include provisions for an augmented leakage l detection system or other license conditions developed during the rulemaking action.
l Applicants for operating licenses seeking to modify design features to take advantage of this rule are required to reflect the revised design in an amendment to the pending FSAR. If the design change modifies design criteria set forth in the PSAR, an amendment to the applicable construction permit may also be necessary. The amendment to the FSAR, and the applica-tion for amendment of the construction permit if necessary, may include O
V provisions for augmented leakage detection or other design or operating features developed during the rulemaking action.
This rulemaking will introduce an inconsistency into the design basis by
- excluding only the dynamic effects of postulated pipe ruptures while still retaining non-mechanistic pipe rupture for emergency core cooling systems, containments, and environmental qualification. The Comission recognizes the need to address whether and to what extent leak-before-break analysis techniques may be used to modify present requirements relating to other features of facility design. However, this is a longer term evaluation.
For the present, the proposed rule allows the removal of plant hardware O
8
[7590-01]
which it is believed negatively affects plant performance and safety, while not affecting emergency core cooling systems, containments, and environmental qualification.
For existing PWRs, considering primary coolant loops only, cost savings of
$186 million and reductions of 34,000 man-rem are estimated for a population of 85 PWRs. These figures did not include savings resulting from redesign of heavy component supports. One licensee seeking to take advantage of the modification of GDC 4 is estimating a per plant cost savings of $20 million and reduced worker exposures of about 2000 man-rem associated solely with a redesign of reactor coolant pump supports.
The above-mentioned value-impacts were realized under the already published Q limited scope amendment to GDC 4. Additional benefits which can be achieved under this broader amendment are discussed below.
For existing BWRs, considering only recirculation loop piping, cost savings of $30 million and reductions of 8,600 man-rem are estimated for a )
l population of 38 plants.
i In existing PWRs and BWRs, public risk is estimated to be insignificantly I impacted, or if credit is taken for improved inservice inspection and enhanced safety, to be reduced by an unquantified amount.
The Commission has not quantified situations in existing plants other than l
those discussed above; however, it is believed that other high energy I piping will also indicate favorable value-impacts.
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[7590-01]
- Value-impacts resulting from this rule are greatest for future plants, where estimated costs can be reduced approximately $100 million per unit.
Of this sum, about $30 million are direct costs and the balance stems from reduced financing costs and improved scheduling. Reduction in worker radiation exposures vary from plant to plant, but are in the range of 300 to 800 man-rem. Public risk was not quantified, but is believed to decrease due to improved effectiveness of inservice inspection and enhanced safety. The above quoted figures are based primarily on the elimination of pipe whip restraints and jet impingement barriers and do not treat other facility changes that could result from this rule.
i
SUMMARY
OF ACCEPTANCE CRITERIA O
V The Commission plans to begin development of a Regulatory Guide and a Standard Review Plan section dealing with the proposed acceptance criteria associated with the Broad Scope modification of GDC 4 after review and evaluation of solicited public comment have been completed.
While NRC acceptance criteria are subject to further elaboration and revision as the results of ongoing studies become available, some details ,
1 on existing staff guidance (based primarily on Chapter 5 of Volume 3, NUREG-1061 dated November 1984) are given below.
The leak-before-break approach should not be considered applicable to fluid system piping, or portions thereof, that operating experience has indicated is particularly susceptible to failure from the effects of corrosion (e.g.,
intergranular stress corrosion cracking), water hamer, creep or low and 10
[7590-01]
high cycle (i.e., thermal, mechanical) fatigue. To show that piping systems are not susceptible to failure from water hammer, creep and corrosion, the extremely low probability criterion must be satisfied. This can be accomplished through investigations of operating history and measures to prevent or mitigate these phenomena. Creep is not a problem in light water reactors, but may occur in other reactor types covered by this rule.
The leak-before-break approach should not be considered applicable if there is a high probability of degradation,or failure of the piping from indirect causes such as fires, missiles, and damage from equipment failures (e.g.,
cranes), and failures of systems or components in close proximity to the pipe. Piping systems, or portions thereof, located in non-Seismic Category O I structures may take advantage of leak-before-break technology when it can be shown that these structures, either as originally designed or through redesign, can resist SSE loads with acceptable margin.
The leak-before-break approach is applicable only in piping systems, or portions thereof, which have been verified to accurately reflect in their 1 analyses the as-built configuration (as opposed to the design configuration).
The leak-before-break approach is limited to piping systems where the material is not susceptible to cleavage-type fracture over the full range of systems operating temperatures.
O Leakage detection systems should be sufficiently reliable, redundant, diverse and sensitive so that a margin no less than 10 on detection of 11
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[7590-01]
< l unidentified leakage from throughwall flaws exists. State of the art leakage detection requirements equivalent to Regulatory Guide 1.45 must be satisfied for all piping within the scope of this rule.
Also, residual welding stresses and cold springing stres;es on a case-by-case basis may require special remedies.
In performing the fracture mechanics evaluations, the following procedure i meets the proposed acceptance criteria.
i Specify the type and magnitude of the loads applied (forces, bending and torsional moments), their source (s) and method of combination.
I 4
O Identify the iocatiencs) at which the highest stresses coincident with poorest material properties occur for base materials, weldments, and safe ends.
1 Identify the types of materials and materials specifications used for base metal, weldments and safe ends, and provide the material properties including apppropriate toughness and tensile data, long-term effects such as thermal aging and other limitations.
] Postulate a flaw at the location (s) where the highest stresses coincident with poorest material properties occur for base materials, weldments and safe ends and that would be permitted by the acceptance criteria of Section XI of the ASME Boiler & Pressure Vessel Code. Demonstrate by fatigue crack growth analysis that the crack will not grow significantly during service.
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Postulate a throughwall flaw at the location (s) above. The size of the
/~T U
flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection capability when the pipes are subjected to normal operating loads. If auxiliary leak detection systems are relied on, they should be described.
Assume that a safe shutdown earthquake (SSE) occurs prior to detection of the leak to demonstrate that the postulated leakage flaw is stable under normal operating plus SSE loads for a long period of time; that is, crack growth, if any, is minimal during an earthquake.
Determine flaw size margin by comparing the selected leakage size flaw to critical size crack. Using nonnal plus SSE loads, demonstrate that there O 4s a mergin of at ieest 2 between the ieekage s4ze fiaw and the criticai size crack to account for the uncertainties inherent in the analyses and leak detection capability.
Determine margin in terms of applied loads by a crack stability analysis.
Demonstrate that the leakage-size cracks will not exprience unstable crack growth even if larger loads (at least dtimes the normal plus SSE loads) are applied. Demonstrate that crack growth is stable and the final crack size is limited such that a double-ended pipe break will not occur.
The piping materials toughness (J-R curves) and tensile (stress-strain l
curves) properties should be determined at temperatures near the upper range of normal plant operation. The test data should demonstrate ductile behavior at these temperatures.
13
- i. [7590-01]
1 l
The J-R curves should be obtained using specimens whose thickness is equal l or greater than that of the pipe wall. The specimen should be large enough
! to provide crack extensions up to an amount consistent with J/T condition i
determined by analysis. Because practical specimen size limitations exist, a
the ability to obtain the desired amount of experimental crack extension may be restricted. In this case, extrapolation techniques may be used. t i
The stress-strain curves should be obtained over the range from the l proportional limit to maximum load.
i
)
The materials tests should be conducted using archival material for the j pipe being evaluated. If archival material is not available, tests should be conducted using specimens from three heats of material having the same i O -teriai specificati . Test -teriai shouid inciude base and weid metais.
l i At least two stress-strain curves and two J-resistance curves should be
\
i developed for each of a minimum of three heats of materials having the same
! material specifications and thermal and fabrication histories as the I
j in-service piping material. If the data are being developed from an i
archival heat of material, a minimum of three stress-strain curves and i.
f three J-resistance curves from that one heat of material is sufficient.
1 i
The tests should be conducted at temperatures near the upper range of normal plant operation. Tests should also be conducted at a lower temperature, which may represent a plant condition where pipe break would 5
i present safety concerns similar to normal operation. These tests are O
4 14
- [7590-01] l l
One J-R toughness on temperature over the temperature range of interest.
curve and one stress-strain curve for one base metal and weld metal are considered adequate to determine temperature dependence.
There are certain limitations that currently preclude generic use of analyses evaluate leak-before-break conditions for limit-load to eliminating pipe restraints. However, limit-load analysis can be used to demonstrate acceptable leak-before-break margins, provided the limit moment is greater than the applied (nonnal operation plus safe shutdown earthquake (SSE)) moment at any location in the pipe run by a factor of at least three.
INVITATION TO COMMENT 1
Comment is invited on the following topics:
A
- 1. Value-impacts associated with this expanded modification to GDC 4, with particular reference to experience with the use of pipe whip ;
' restraints and jet impingement shields near nuclear reactor piping.
(The value-impact analysis prepared by Lawrence Livermore National ,
Laboratory is available for inspection and copying for a fee in the NRC Public Document Room,1717 H Street NW, Washington, D.C.)
i
- 2. The scope of piping which could or should be affected, supported by technical justifications.
15
[7590-01)
- 3. The decision to limit impacts of this modification of GDC 4 to only dynamic effects associated with pipe rupture.
- 4. The acceptance criteria which the Comission proposes to use to evaluate whether leak-before-break technology is applicable to specific situations.
i
- 5. Acceptable allowables for pipe-connected component supports which 1 would provide adequate assurance that component support failure would not be a source of the pipe rupture loads being eliminated from the design basis.
AVAILABILITY OF DOCUMENTS O
- 1. Copies of NUREG-1061, Volume 3, may be purchased by calling (202) 275-2060 or (202) 275-2171 or by writing to the Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082 Washington, D. C., 20013-7082, or purchased from the National Technical Information Service, Department of Comerce, 5285 Port Royal Road, Springfield, VA 22161.
- 2. Copies of NUREG/CR-4263, may be purchased by calling (202) 275-2060 or (202) 275-2171 or by writing to the Superintendent of Documents, U.S.
Government Printing Office, Post Office Box 37082, Washington, D.C.,
20013-7082, or purchased from the National Technical Information Service, Deparment of Comerce, 5285 Port Royal Road, Springfield, VA 22161.
16
- [7590-01]
FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY
(
The Commission has determined under the National Environment Policy Act of 1969, as amended, and the Commission's regulations in Subpart gf 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required. Although the removal of certain plant hardware could result, this will not alter the environmental impact of the licensed activities as set out in the Final Environmental Impact Statement for each facility. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room,1717 H Street, j NW, Washington, DC. Single copies of the environmental assessment and the O <iadi"9 or ao sia"$<icaat 4 Pact re vati bie <ro aoh" ^ o Briea. 'rrice Nuclear Regulatory Commission, of Nuclear Regulatory Research, U.S.
Washington,DC20555, telephone (301)443-7854.
PAPERWORK REDUCTION ACT STATEMENT This proposed rule does not contain a new or amended infonnation collection requirement subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501
, et seq.). Existing information collection requirements under 10 CFR Part 50 were approved by the Office of Management and Budget approval number 3150-0011.
O 17
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[7590-01)
REGULATORY ANALYSIS The Comission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the
' alternatives considered by the Commission. The draft analysis is available for inspection in the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the analysis may be obtained from John A.
O'Brien, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, telephone (301) 443-7854. The Comission requests public coment on the draft regulatory analysis. Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES heading.
O- BACKFIT RULE A backfit analysis (see 10 CFR 50.109) has been prepared and is available for inspection in the NRC Public Document Room, 1717 H Street NW, Washington, D.C. The backfit analysis considers all of the factors listed in 10 CFR 50.109(c), and shows that substantial benefits, both safety and economic, can be gained by application of leak-before-break technology in appropriate circumstances. A fonnal finding of " substantial increase in public health and safety" [see 10 CFR 50.109(a)(3)] has not been made because this amendment is permissive in nature and does not impose any new or changed requirements on licensees. Single copies of the backfit analysis may be obtained from John A. O'Brien, Office of Nuclear Regulatory Nuclear Regulatory Comission, Washington, DC 20555, O Research, U.S.
telephone (301)443-7854.
18
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[7590-01]
O REGULATORY FLEX 1BILITY ACT CERTIFICATION As required by the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)),
the Commission certifies that this rule, if adopted, will not have a significant economic impact on a substantial number of smay. entities.
This rule affects only the licensing and operation of nuclear power plants.
The companies that own these plants do not fall within the scope of the d definitions of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.
LIST OF SUBJECTS IN 10 CFR PART 50 O
Antitrust, Classified infomation, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, i Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.
- For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the 1
following amendments to 10 CFR Part 50.
PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES O 1. The authority citation for Part 50 continues to read as follows:
! 19 I
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[7590-01]
h V
AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, asamended(42U.S.C.5841,5842, 5846),unlessotherwisenoted.
1 Section 50.7 also issued under Pub. L.95-601, sec.10, 92 Stat. 2951 a
(42 U.S.C. 5851). Sections 50.57(d), 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2133, 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec.184, 68 Stat. 954, as amended (42 U.S.C.
2234). Sections 50.100 - 50.102 also issued under sec. 186, 68 Stat. 955 (42U.S.C.2236).
For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.
2273), il 50.10(a), (b), and (c), 50.44,50.46,50.48,50.54,and50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b));
il 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(i)); and li 50.55(e), 50.59(b), 50.70, 50.71, 50.72, 50.73, and 50.78 are issued under sec. 1610, 68 Stat. 950, as ,
amended (42 U.S.C. 2201(o)). l
- 2. In Appendix A, General Design Criterion 4 is revised to read as follows:
- O APPENDIX A - GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS i
i 20
[7590-01]
CRITERIA I. Overall Requirements i
Criterion 4 - Environmental and dynamic effects design bases.
Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging i
fluids, that may result from equipment failures and from events and O conditions outside the nuciear power unit. so ever, dvnamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. These analyses must include, as a minimum, a deterministic fracture mechanics evaluation of the piping, and an evaluation of corrosion, water hanner, creep, fatigue, leakage detection and indirect sources of pipe rupture.
Dated at Washington, D.C. this day of 1986.
For the Nuclear Regulatory Commission.
Samuel J. Chilk, Secretary of the Commission.
BACKFIT ANALYSIS Broad Scope Modification to General Design Criteria 4
- 1. Statement of the specific objectives that the proposed backfit is designed to achieve.
This rulemaking offers licensees and applicants new options not pre-viously available while imposing no additional requirements. Howev-er, physical features at operating plants, plants under construction and in future designs can be impacted when the conditions of the rule are satisfied.
Specifically, licensees and applicants can at their own choosing de-cide to undertake investigations to demonstrate that particular pip-ing systems meet NRC acceptance criteria for applying leak-before-break technology. Successful demonstration will allow elimination of dynamic effects associated with pipe rupture from the design basis while not influencing requirements for environmental t qualification of electrical and mechanical equipment, containment design or emergency core cooling system performance. When dynamic effects are eliminated, certain plant hardware (such as pipe whip i restraints and jet impingement shields) need not be installed and other related changes can take place. In particular, internals and supports of pipe connected components, such as steam generators, and reactor coolant pumps need not be designed for the dynamic effects associated with pipe rupture. Pressurizations in cavities, compart-ments and subcompartments which are not part of the containment sys-tem can also be eliminated under this rulemaking action. The removal of pipe whip restraints improves safety by eliminating potential and unforeseen contact between piping and pipe whip restraints (which would introduce additional stresses by restricting thermal growth).
Moreover, the effectiveness of inservice inspection is enhanced, worker radiation exposures are reduced and construction economics are O improved.
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- 2. -General description of the activity that would be required by the
[: licensee or applicant in order to complete the backfit.
! In order for licensees or applicants to take advantage of leak-before-break technology they must provide sufficient information .
) to satisfy stringent NRC leak-before-break acceptance criteria. For ,
! example, materials test data must be available to support the j leak-before-break evaluation. The acceptance criteria require leak-i age detection systems which are diverse, reliable, redundant and sen- i sitive such that a margin of a factor of 10 exists on undetected i 3
i leakage from the crack assumed in the fracture mechanics analyses.
Applicants or licensees electing to implement leak-before-break (
! technology may need to upgrade or improve existing leakage detection j systems, although it is possible that some plants may have adequate l j leakage detection systems already. Modifications of the licensed [
plant design of operating plants may involve an unreviewed safety :
i question under 10 CFR 50.59. Where it is determined that an lO j
unreviewed safety auestion is invoived, iicensees of operatins piants desiring to make modifications should submit a license amendment for l
} NRC approval in accordance with revised GDC-4. A simple removal of j pipe whip restraints and jet impingement barriers would not involve i an unreviewed safety question. However, modification to component internals and supports would involve an unreviewed safety question. ;
fj Applicants for operating licenses seeking to modify design features j to take advantage of revised GDC-4 are required to reflect the ;
j revised design in an amendment to the pending FSAR. If the design l 4 change modifies design criteria set forth in the PSAR, an amendment to the applicable construction permit may also be necessary. j i
l
! 3. Potential change in risk to the public from the accidental offsite l l release of radioactive materials. l 1
i Best estimate increases in public risks are a total of .003 man-rem l from a population of 85 PWR primary loops and .17 man-rem from 38 I
i !
- z !
i
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BWR recirculation loops. When all piping in the plant is considered, j public risk is not likely to increase by more than a few man-rem.
} These best estimates consider only the potential for protective i devices to mitigate accidents. Thus, according to these best
! estimate calculations, when the accident mitigation potential is removed, public risk increases. The contractor (LLNL) that prepared the Regulatory Analysis for this rulemaking noted that:
t i
"Many experts believe, however, that the proposed Broad Scope rule !
change could conceivably reduce public risk, arguing that pipe whip
- restraints and jet impingement barriers may actually decrease the reliability of piping systems by limiting access to pipe welds, therefore reducing the effectiveness of in-service inspection, and by i potentially restricting the movement of piping during routine opera-4 tion and thus increasing pipe stresses caused by restraint of thermal j expansion. This latter situation could occur, for example, if pipe j whip restraints (which are designed to acconunodate thennal expansion
{Q displacements measured during preoperational plant testing) were to be installed incorrectly. In this value-impact assessment, however, l
i we have assumed that restraints are installed as designed and have I i i j not attempted to quantify any effect that avoidance of possible pipe l l binding might have on public risk. Similarly, we have not attempted j to take credit for increased piping reliability due to improved j effectiveness of in-service inspection." (UCID-20397) l l 4. Potential impact on radiological exposure to facility employees.
i i i
For the population of 85 PWRs (considering only primary coolant j loops) and the population of 38 BWRs (considering only the l recirculationloop),occupationalexposuresarereducedby34,000and j 8,600 man-rem respectively. When the scope of piping is expanded to j other piping inside the containment, these best estimates could be increased severalfold, l
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h 5. Installation and continuing costs associated with the backfit, in-cluding the cost of facility down times for the cost of construction delay.
For future plant designs, cost savings of approximately $100 million per plant can be achieved, of which approximately $30 million is di-
- rect cost and the remainder is due to improved construction schedule For the 85 PWRs operating or under con-and reduced financing costs.
struction and considering only the primary coolant loops, cost sav-ings of about $200 million result while for the 38 BWRs operating or under construction and considering only the reactor coolant loop, cost savings of about $30 million result. Additional benefit is proportional to the extent that LBB is applicable to other piping.
- 6. The potential safety impact of changes in plant or operational com-plexity including the effect on other proposed and existing regulato-ry requirements.
t The proposed rule is calculated to improve safety and increase the effectiveness of inservice inspection, reduce worker radiation expo-sures and positively impact construction and maintenance economics.
This rulemaking allows the removal of counterproductive hardware from the plant and leads to significantly less complexity in plant op-erations. Fire protection is enhanced since movement about the plant is improved. Thermal efficiency is also improved since insulation may now be added to sections of high energy piping which previously, because of pipe whip restraints, had no pipe insulation.
- 7. The estimate resource burden on the NRC associated with the proposed backfit; and the availability of such resources.
1 It is estimated that ten to twenty NRC man years of effort to review
! industry initiatives and additional research over the next several years are needed to respond to actions taken under the Broad S: ope 4
GDC-4 Rule. In view of Gramm-Rudman constraints, prioitization within ELD, NRR and RES would need to be adjusted to make the resources available.
- 8. The potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit.
Because of existing problems with stress corrosion cracking in BWRs, the Broad Scope GDC-4 amendment is expecteo to have far greater impacts on PWRs than BWRs. Age is not a factor th implementing this rule. Value-impacts are more in future plants and less in operating plants.
Any backfits under the Broad Scope GDC-4 modification are final.
l 1
O S
NRC PROPOSES TO CHANGE REQUIREMENTS FOR REACTOR PIPING The Nuclear Regulatory Comission is proposing amendments to its
~
regulation which has led to the use of numerous restraints and jet impingement shields for cooling water pipes in licensed nuclear power reactors. This new amendment expands the scope of a similar more limited
! prior amendment.
The existing regulation is based on the assumption that cooling water pipes could rupture suddenly and requires that nuclear power plants be f
i designed and built to assure that they could be shut down safely in the i event of a sudden pipe rupture.
. l More recent technological developments, however, indicate that piping used in nuclear plants will leak at a detectable rate before rupturing. This i
j
} means that leakage from flawed pipes can be detected and the pipes f repaired before a sudden rupture takes place, making the use of many restraints and jet impingement shields to protect against the forces which i
j would be associated with a pipe rupture unnecessary.
!O Limiting the use of these structures also would reduce radiation exposures to workers who have to remove and replace them when piping inspections are f performed and would avoid safety hazards associated with improper
! placement of pipe restraints.
As proposed, the amendments would not make any changes to the Comission's existing requirements dealing with loss of cooling water, pressure l
! increases in reactor containment buildings or protection against i
earthquakes. This new technical approach was recently approved for a certain portion of piping directly connected to one reactor type, and is l now being applied to other piping as well as other types of light water reactors.
)
! Written coments on the proposed amendment to Part 50 of the NRC's regulationsshouldbereceivedby(date). They should be addressed to the l
) Secretary of the Comission Nuclear Regulatory Comission Washington, iO 0.C. 20sss. attention: Oocketin and Service eranch.
i <
i 1
/ %, UNITED STATES 8 g NUCLEAR REGULATORY COMMISSION g a wAsHawaTom, p. c.sosss I
l The Honorable Morris K. Udall, Chairman l Subcomittee on Energy and the Environment _ l Comittee on Interior and Insular Affairs United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
Enclosed for the information of the Subcommittee are copies of a public announcement and a proposed amendment to Title 10 of the Code of Federal Regulations which is to be published in the Federal Register.
The Nuclear Regulatory Connission is proposing to amend its regulations by modifying General Design Criterion 4 in Appendix A of 10 CFR part 50 to allow demonstration of piping integrity by analysis to serve as a basis for excluding consideration of dynamic effects associated with certain pipe ruptures. The modification will permit the selective removal of pipe whip restraints and jet impingement shields from operating plants, plants under construction and future plant designs, but will not impact other design requirements, as for example, emergency core cooling system (ECCS) perfomance and containment design.
o The Commission had^.previously undertaken a smaller scope rulemaking applicable to the iirimary loops of pressurized water reactors. This present action is of much broader scope in that it includes all piping in all light water reactors that meet specific acceptance criteria.
The Comission is issuing the proposed amendment for a sixty-day public coment period.
Sincerely, Robert B. Minogue, Director Office of Nuclear Regulatory Research
Enclosures:
- 1. Public Announcement
- 2. Federal Register Notice cc: Rep. Manuel Lujan IDENTICAL LETTERS SENT TO THOSE ON ATTACHED LIST O
r h The Honorable Alan Simpson, Chairwan Subcommittee on Nuclear Regulation Comittee on Environment and Public Works i
1 United States Senate Washington, DC 20510 ,
cc: Sen. Gary Hart The Honorable Edward J. Markey, Chairman Subcomittee on Energy Conservation and Power Comittee on Energy and Comerce United States House of Representatives Washington, DC 20515 cc: Rep. Carlos Moorhead O
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ENVIRONMENTAL ASSESSMENT I FOR PROPOSED REGULATION TO BROADEN , SCOPE OF
> ^ GENERAL DESIGN CRITERION 4 MODIFICATIONS
}i Identification of the Proposed Action f
The Nuclear Regulatory Comission is proposing to modify General Design Criterion 4 of Appendix A, 10 CFR Part 50 to allow demonstration of piping integrity by analyses to serve as a basis f6r excluding consideration of dynamic effects associated with poht01ated pipe ruptures. The 4
modification will permit the selective , removal of pipe whip restraints and jet impingement barriers and other related changes in operating plants, i plants under construction, and future plant, designs. This present action O exRand, the scope of a simiiar more iimited amendment which was directed at primary loops of pressurized water reactors.
i
! The Need for the Proposed Action l Advances in technology have led to the acceptance by the NRC staff of procedures that estimate the likelihood of ruptures in nuclear reactor i
l piping. However, General Design Criterion 4 (GDC 4) does not allow use of i this new technical approach except by exemption granted pursuant to 10 CFR 50.12. Rulemaking is therefore needed to accomodate this engineering advance.
i i Prior to the last few years, there was no sound technical basis for excluding certain pipe ruptures from the design basis. Now it is clear
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i that it is possible to' defend the exclusion of pressurized water reactor i primary loop double-ended guillotine pipe ruptures, and that the scope may i be extended to other piping, including piping in boiling water reactors.
! Rulemaking action will promote investigations to determine which other situations will permit the removal of pipe whip restraints and jet
! impingement barriers. Acceptance criteria for generally applying these 1
l results pertaining to' leak-before-break have been published by the NRC i
staff in " Report of the U.S. Nuclear Regulatory Commission Piping Review l _
Comittee", NUREG-1061. Volume 3. .end are being proposed by the American Nuclear Society in ANS-58.2 entitled " Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture."
l In summary, the requirements of GDC 4 have led to a situation where i O protective devices have been added to nuciear ,ower ,iants to forestaii events which are now regarded as extremely unlikely. These protective l
j devices reduce safety and increase worker radiation exposures. A need exists to allow exclusion other than by exemption from compliance with General Design Criterion 4 requirements when supported by acceptable analyses.
Alternatives Considered I
i Three alternatives to the proposed rule were evaluated as indicated below:
- 1. Maintain the status quo l
l This effectively would continue to require the placement of pipe whip 1
restraints and jet impingement barriers on all piping except the primary 1
i
b 3
h main loops of PLRs. This alternative is rejected because analyses have shown that safety could actually be enhanced since misalignment or not maintaining tolerances when installing or reinstalling pipe whip restraints actually increases the probability of pipe rupture. Moreover, substantial cost savings can be realized when these protective devices are removed and total worker man-rem exposures will be reduced,significantly.
- 2. Reinterpret the existing text of GDC 4 )
For more than fifteen years the staff has interpreted GDC 4 to >
require the placement of pipe whip restraints and jet impingement barriers near nuclear reactor piping. Because of legal restraints on issuing I
permanent exemptions, this alternative is rejected. Rulemaking is necessary to justify the departure from long-standing past practices.
- 3. Use Eiemptions to Accomplish the Removal of Fipe Whip Restraints and Jet Impingement Barriers While rulemaking has recently been implemented for the primary loops of PWRs, the use of plant specific exemptions to the regulations on a
> ' system unique basis entails significant allocation of NRC resources. The use of repeated GDC 4 exemptions amounts to an amendment to a fundamental NRC rule in the absence of rulemaking procedurks, leading to potential legal difficulties. For these reasons, this alternative is also rejected.
4,.
l
4 Based on this evaluation, the staff recommends that the NRC proceed pJ with the proposed rulemaking because:
- 1. It fimly secures the legal basis for staff actions.
- 2. It removes impediments to the application of new technology in the licensing arena, thereby allowing the realization of improved safety, lower costs and reduced worker radiation exposures.
Environmental Impacts of the Proposed Action The proposed action would not significantly affect the quality of the human environment. The rule may result in the removal of certain plant hardware; however, this will not alter the environmental impact of the O iiceesed activities as set oet 4n 18e rinai Eevironmeetai Impact statemeet for each facility. The proposed action would substantially reduce occupational radiation exposures received by workers undertaking inservice inspection, and may actually reduce the probability of core melt accidents when potential errors in installation or reinstallation of pipe whip restraints are considered. The staff has already granted exemptions allowing some utilities to remove pipe whip restraints. In sumary, the proposed action would have no measurable negative environmental impacts.
Agencies and Persons Consulted The proposed action has been reviewed by the NRC staff and two NRC con-tractors (Lawrence Livermore National Laboratory and Battelle Pacific Q
V i
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5 O aert8 est ' 8er teries). ^ arenosee reie sii de ve8iis8e8 e pediic coments will be requested.
Findings of No Significant Impact The Comission has determined not to prepare an environmental impact statement for the proposed action because it would not have a significant effect on the quality of the human environment; the impact will be a benefical, not a deleterious, one.
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