ML20137P585
| ML20137P585 | |
| Person / Time | |
|---|---|
| Issue date: | 01/30/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1485, NUDOCS 8602050261 | |
| Download: ML20137P585 (251) | |
Text
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GCASNM Tr o UNITED STATES NUCLEAR REGULATORY COMMISSION ORIGINAL IN THE MATTER OF: DOCKET NO:
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE ON ADVANCED REACTORS ,
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LOCATION:. WASHINGTON, D. C. PAGES: 1 - 157 DATE: THURSDAY, JANUARY 30 , 1986
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CR25685.0 DAV/dnw 1 UNITED STATEb OF AMERICA Ns) 2 NUCLEAR REGULATORY COMMISSION 3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4 SUBCOMMITTEE ON ADVANCED REACTORS 5
6 Nuclear Regulatory Commission Room 1046 7 1717 H Street, N.W.
Washington, D. C.
8 Thursday, January 30, 1986 9
The meeting of the subcommittee convened at 8 : 35 a.m. ,
10 Dr. Max W. Carbon presiding.
11 ACRS MEMBERS PRESENT:
12 7- DR. MAX W. CARBON
(,) 13 DR. CARSON MARK DR. CHESTER P. SIESS 14 15 .
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i 18 l 19 i 20 21 22 l
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PUBLIC NOTICE BY THE
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UNITED STATES NUCLEAR REGULATORY COMMISSIONERS' ADVISORY COMMITTEE ON REACT'OR SAFEGUARDS THURSDAY, JANUARY 30, 1986 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 inacc'uracies of statement or data contained in this transcript.
O
a 6850 01 01 2 1 DAVbw 1 PROCEEDINGS 2 DR. CARBON: The meeting will now come to order.
3 This is a meeting of the Advisory Committee on 4 Reactor Safeguards, Subcommittee on Advanced Reactors.
5 My name is Carbon. The other ACRS members here 6 today are Dr. flark and Chet Siess, who should be here 7 momentarily.
8 The purpose of the meeting is to review the HTGR 9 design that was submitted to NRR by DOE and to determine the 10 basis for selection of the design basis events.
11 Dr. El-Zeftway is the cognizant ACRS Staff member for the 12 meeting.
13 Rules for participation in today's meetings have O' 14 been announced in the Federal Register.
15 A transcript is being kept. It's requested that 16 each speaker identify himself or herself, use the mike, and 17 so on.
18 We received no written statements from members of 19 the public and no requests for time to ma:<e statements from 20 members of the public.
21 I have just a couple of comments to make before 22 we begin.
23 This meeting is a consequence of a letter from 24 Carl Neal to Ray Fraley on November 26, in which NRR and DOE 25 asked to interact with the Advance Reactor Subcommittee on ACE-FEDERAL REPORTERS, INC.
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6850 01 02 3 1 DAVbw 1 HTGRs and LMR designs. At the present time, at least the eg
-) 2 ACRS is planning to go ahead with this. We are suffering 3 like everyone else from some budget cuts, and we're going to 4 have to make some decisions as to what we do and what we 5 don't do. But for the present, at least, plans are to 6 continue the interactions with your two groups. It will 7 tentatively be our intent to pass on comments, obviously, 8 during the discussions on days like this. We may very well 9 forward written comments after discussion with the full 10 committee, but this is not determined at the present time, 11 ,
We have today's meetings scheduled on HTGR. Then 12 there is a meeting scheduled for late February on the LMR.
13 One other comment. We're planning to charge
(_) 14 right through to 2:00 o' clock and go without lunch. If this 15 gives anyone any problems, let us know. It will be helpful 16 on plan reservations for us, and so on.
17 Carson, do you have any comments?
18 DR. MARK: Not really. But generally, perhaps, 19 the proposals for advanced reactors, of which this is one, 20 and an attractive looking, in what degree does our 21 participation affect the time scale? The time scale is 22 really affected by when somebody gets down to business and 23 says we want one. At that point, we need to come in, 24 indeed. Right now we're interested, indeed, in hearing 25 about the design thoughts which people have, what is it we O
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6850 01 03 4 1 DAVbw 1 can do about advancing or promoting or accelerating or n
kJ 2 simplifying the process towards such a thing happening.
3 That's the kind of question I have in mind.
4 DR. CARBON: I think I can provide a partial 5 answer at least.
6 We've encouraged NRC and NRR to work with DOE, 7 early in the stage of design and so on. I know I personally 8 am a strong believer that safety ought to come in early and 9 be built in as part of the overall design aspects right from 10 day one. I think it ends in a much better product, both 11 from the standpoint of the economic side, the operational 12 side, but also from the safety side, and I think this is 13 part of the process of not only NRR interacting with DOE, 14 but also DOE and NRR asking us for our comments.
15 If we were to have strong objections to what they 16 were doing and these didn't show up till way late in the 17 i design stage, as a bare minimum, it would certainly lead to f
18 some bad inefficiencies and repetitious design effort and 19 that sort of thing.
20 DR. MARK: Well, I'm going along on that 21 hypothesis too.
22 DR. CARBON: Chet?
23 DR. SIESS: Max, I guess I've been looking at 24 this in the context of what we've done in the past, 25 reviewing a concept. Remember, we reviewed the concept of 7-V ACE-FEDERAL REPORTERS, INC.
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6850 01 04 5 2 DAVbw -1 upper head injection, ice condensors, things of that sort
\ 2 and picked up on certain things.
3 If you look at the conceptual design for the 4 small HTGR, I see two issues that the ACRS could easily be 5 involved in at some point which are crucial to the concept, 6 and indeed, could be addressed.
7 Now there may be others. One of them is the 8 question of whether there's a containment or not. This fits 9 into the philosophy of what doses are, how do you accept 10 extreme accidents and so forth.
11 I mention that, because that was an issue of the 12 only other gas-cooled that we've reviewed to completion --
13 Fort St. Vrain.
14 When we say other gas-cooled, they had 15 containments. There was a strong recommendation in the ACRS 16 letter that the next one, a bigger one, ought to have a 17 containment, and I think whether the design has a 18 containment or not is a major point, because that affects 19 the economic viability of it, and it's going to affect some 20 of the conceptual approach.
21 I think another aspect of the concept that the 22 ACRS ought to look at and agree with somebody on is the 23 separation of a safety-related system from a 24 nonsafety-related system. I think that accepting the idea 25 that something of this sort was not safety-related, and this ACE-FEDERAL REPORTERS, INC.
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6850 01 05 6 1 DAVbw 1 one is, again is important to the economic viability and O)
\s 2 safety viability of the design.
3 I'd hate to try to go through the definitions of 4 " safety-related" or "important to safety," which I'm not 5 sure are settled, but there are things that we didn't hinbk 6 were safety-related on lightwater reactors at one time, that 7 we not think are safety-related, the outstanding example 8 being aux feedwater. And now we're looking at all sorts of 9 things that could be an initiator of a transient, an 10 initiator of an accident, or a challenge that isn't 11 safety-related that could be a challenge. I think of 12 safety-related as only those things that are required to 13 mitigate an accident and so forth. But again, that
[)
14 separation, I think, is conceptual.
15 I think it's important to the design, the 16 viability of the design. And I think it's the sort of thing 17 that the ACRS would have an opinion on, and they're not 18 going to have a good opinion on it, until they understand a 19 lot about the design.
20 So those are two things. There was a third one I 21 had in mind. I guess the third point isn't all that clear, 22 from what I heard. It may get cleared up today. We're 23 still having a problem like water reactors in separating 24 design basis accidents from severe accidents. The severe 25 accident is being treated on a probabilistic basis. Design ACE-FEDERAL REPORTERS, INC.
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6850 01 06 7 1 DAVbw 1 basis accidents being a deterministic type of thing. And 2 except for the severe accident policy statement, we don't i
3 have very much on severe accidents. I guess you could say, 4 even with the severe accident policy statement, we don't I
- 5 have very much, because nobody knows how to relate it to a i
j 6 safety goal, because we don't have a safety goal. ,
! 7 So those are the kinds of things I think the 8 ACRS could address, although the ACRS can probably never 1
6 9 comment itself to what a future ACRS might do.
10 You get my point. A review of the concept and l
j 11 trying to find those features of the concept that are 12 important, not the details of it, not whether it's a
, 13 concrete vessel, a steel vessel or whatver.
( 14 DR. MARK: Chet, I think you put things i
15 beautifully. Just r ig ht. That's where we are and where we j 16 need to concern ourselves.
1 17 DR. SIESS: You know, there are going to be 18 things like the purely passive concept, the walk away, the
- 19 extremely low doses, and I guess if everybody could be
- 20 convinced that the consequences are going to be small, I i
l 21 don't care what you do, you can shift gears and get into a 22 different domain of thinking.
23 Now whether anybody can get to that stage or the 24 conceptual review, I don't know, but anything that will keep 25 moving in that direction would help.
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l 6850 01 07 8 1 DAVbw 1 DR. CARBON: Fine. I too share the view that
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k 2 you've put it very, very well, and also, share the technical 3 views.
4 Let's go ahead then and call on Tom King.
5 (Slide.)
6 MR. KING: My name is Tom King, with the Office 7 of Nuclear Reactor Regulation. I'm a Section Leader in the 8 Safety Program Evaluation Branch in the Division of Safety 9 Review and Oversight. My section has the responsibility for 10 the advanced reactor reviews and interactions with DOE on 11 both thge HTGR and the LMR designs.
12 I think Dr. Carbon summed it up well this 13 morning. The purpose of these interactions is to work early 7_s
\ )
14 on in the design process to factor in safety and licensing 15 considerations into the design, so that we don't get further 16 downstream and have to work these issues out and cause a 17 rework of the design or patchwork-type fixes. We're trying 18 to incorporate safety into the design right from day one.
19 These interactions are consistent with the 20 Advanced Reactor Policy Statement which not only calls for 21 the Staff to interact early but also request ACRS 22 involvement in these interactions.
23 The purpose of today's meeting primarily is to 24 acquaint ACRS Subcommittee members with the HTGR design.
25 DOE and its contractors will be doing most of the talking ACE. FEDERAL REPORTERS, INC.
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t 6850 01 08 9 1 DAVbw 1 this morning. I'm just going to give a quick overview of l
( 2 where our review plans stand at the present time.
3 What we've agreed to with DOE is to review the 4 conceptual design of the HTGR. This is a process that will l
5 take place approximately over the next two years, with the 6 primary purpose being to develop and agree on criteria for 7 the design, to assess the potential of that design for 8 satisfying these criteria, to assess the R&D programs 9 supporting that design. And at the end of the two-year 10 process, to issue an SER and what we call a licensability 11 statement.
12 At the end of this two-year period, we're 13 planning, or we hope to be in a position then, to identify O 14 additional steps. Research, for example, that need to be 15 taken for the NRC to be ready to process and HTGR 16 application for either standard plant review or an actual 17 commercial plant.
18 DR. MARK: What authority -- I'm not sure that's 19 the right word. What authority, let me say it again, do you 20 have -- supposing you come to the conclusion, which is a 21 possible conclusion, that an HTGR, to be specific, need or 22 need not have high level containment. What authority do you 23 have then to represent NRR at the agency on that aspect of 24 things?
25 MR. KING: When we would issue an SER, it would ACE-FEDERAL REPORTERS, INC.
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6850 01 09 10 1 DAVbw 1 be NRR's position on what we think of the design, which 2 features we don't like or do like. We would certainly 3 factor in ACRS comments, thoughts, positions on that, but it ,
4 would not have the force of law like an actual license 5 would, of DOE or the contractors could.
6 DR. MARK: It would still be subject to 7 potentially endless debate, whether this was the correct 8 view or not.
9 MR. KING: Potentially, it could. It could go 10 on, but the idea is to try to get a consensus early on, a 11 staff consensus and ACRS consensus and a designer 12 consensus.
13 DR. MARK: That's certainly among the questions O 14 we're going to have to tangle with, I would suggest before 15 very long, perhaps even today.
16 DR. CARBON: A couple of questions. One that i 17 sort of follows Carson's. How much interaction will you 18 have, does your group have with the rest of NRR? When we 19 end up here, you're in effect.saying, well, tentatively NRR 20 feels so-and-so about this. But you say will depend to a 21 considerable part on whether it's just a small group of 22 people, or whether a large part of NRR has had some sort of 23 look or interaction here.
l 24 MR. KING: In the current organization, the new 25 organization at NRR, we will be drawing upon a lot of people ACE-FEDERAL REPORTERS, INC. ;
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, 1 DAVbw 1 in Dr. Spiess' division, the Safety and Oversight Division.
2 DR. CARBON: You are in that? l 3 MR. KING: I am in that Division. My particular :
4 branch will have people involved. We have approximately 10 5 or 12 other people outside of my branch that will be 6 involved. We're going to use NMSS for safeguards and 7 security review. We will probably go to other people in NRR 4
i 8 where they have expertise in specific areas like the fire 9 protection expert is not in. If he's in another division, 10 we will try to use him, but primarily, the review will be i s
11 done in the SRO. They're assigned that responsibility in 12 the new organization.
. 13 DR. CARBON: But is it a fair statement to
( 14 believe that a good-sized segment of NRR will have been 15 exposed to the thoughts and concepts in this and will have 16 given at least a little preliminary thought to it.
17 MR. KING: A fair-sized segment of DSRO will have 18 been exposed to it. The other divisions are vendor 19 divisions, and there will be a limited number of individuals 20 in those divisions who will have been exposed to it.
21 DR. CARBON: Will Spiess have reviewed it?
4 22 MR. KING: Yes.
23 DR. CARBON: How much involvement will Denton i 24 have in this? How much review will he do?
25 MR. KING: I can' t speak for Mr. Denton, but our ACE-FEDERAL REPORTERS, INC. ;
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6850 01 11 12 !
1 DAVbw 1 intention is to issue certainly things like the I
2 licensability letter over his signature.
3 In fact, as we go along, I think a lot of the 4 Staff work and responses to the DOE on the various key 5 issues that we're going to be looking at, our plan is to 6 send those out under Dr. Spiess' s igna tu re . If there's a 7 particular key one, for example, whether there is a I
8 containment or not a containment, I think that kind of issue 9 may get elevated to Denton's level. But we'll have to see i
10 what the issue is and make that decision based on that.
11 We don't have a predetermined plan at this point 12 as to who is going to sign out what, but it's going to be 13 the Spiess-Denton level, depending on what the issue is.
O 14 Our correspondence deals with that.
15 DR. CARBON: The second question refers to the i
16 last bullet there on the SER and licensability statement.
I 17 Could you pin down what that licensability statement will do 18 or say or mean, a little bit more?
19 MR. KING: Our intent at this point is for it to 20 be a summary of what the major features are of that design, 21 the major characteristics of that design we'd like to see to 22 make it licensable. In other words, it will concentrate on 23 those area that maybe we've had a problem with, or we'd like 24 to see some change in.
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6850 02 01 13 1 DAVbur 1 Maybe the best way to look at it is a summary of
,s
- 2 what comes out of the SER, look at the whole design 3 throughout.
4 There will be things that we agree with. There 5 will be things we have problems with.
6 The licensibility statement in the division will 7 be sort of a summary. Here are the major things that we 8 would like to see changed to make this design licensable.
9 DR. CARBON: Will the thrust of it be to say, for 10 example, it is our opinion that this is on a permanent basis 11 and it could probably readily move forward and expect to 12 receive a license somewhere down the road and we only have
,_ 13 problems with two or three particular things and we should b 14 change those, or is it perhaps to be more simply a 15 statement: the only suggestions we can make are these.
16 What is the SER going to say? It can't be 17 final?
18 MR. KING: No. We are dealing with a conceptual .
19 design. It is not going to have a lot of detail in many 20 areas.
21 We are not going to get into, for example, 22 details of analysis codes and validation of those codes.
l l 23 But I think in a broad sense we are going to look at things l
24 like what codes we are going to say for an actual l
l l fs 25 application. You are going to have to demonstrate how you y
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4 6850 02 02 14 1 1 DAVbur 1 validate this code.
! 2 If we see areas in the R&D program where they 3 need additional fuels data to support their fuels, 4 materials data to support their reactor vessel, we are going 1 5 to try to identify those, and the licensibility letter is J
6 going to summarize those and say here are the key areas 7 where we feel additional work needs to be done; some changes 8 need to be made to make this design licensable in the future 9 for an actual application.
10 We are trying to give some early indication of i
11 where we see some weakness and see some additional 12 information required.
l l 13 DR. SIESS: Is it your intention at some stage 5
l-14 between now and licensing to require a PRA? If so, at what 15 stage? If not, why not?
16 MR. KING: The PRA is already planned. You will 17 see it on the next viewgraph at the conceptual stage. It is 18 scheduled to be submitted to us in September of '86.
19 DR. SIESS: That is just a schedule.
20 How helpful would a PRA be in addressing the two 21 issues I have raised -- containment and separation of 22 safety-related and nonsafety-related?
i 23 MR. KING: I think for the.latter it would be
! 24 very useful, identifying what are the initiators, the risks, 25 the probabilities of those occurring, to see how much
, (3)
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6850 02 03 15 1 DAVbur 1 contribution to risk the balance of plant or the nonsafety 2 grade portions of the plant really contribute.
3 DR. SIESS: Why not the first one?
4 MR. KING: Possibly for the first one, also.
4 5 DR. SIESS: I mean, the PRA was conceived to 6 address somewhere a worst case accident, no matter how low ,
. 7 the probability, and if the consequences of a worst case 8 accident at any probability were sufficiently low, you would i 9 have a pretty good argument on the containment issue.
10 MR. KING: Yes, you would. In fact, if you do, 3
- 11 say, a cost-benefit type analysis of adding containment 12 versus not adding a containment, you are going to need PRA I
, 13 type work to evaluate the consequences, to do that kind of O 14 analysis.
15 So on both points it would almost be essential to 1 16 have a PRA.
17 DR. CARBON: Move on.
j 18 (Slide.)
19 MR. KING: This is just a quick table that
, 20 summarizes the major steps over the next two years in the 21 interactions that we are planning to have. I won' t go 22 through everything, just point out a couple of things.
23 We have already had a number of briefings dealing 24 with the licensing approach, top level criteria, selection 25 of licensing basis, defense design basis events, that kind I
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6850 02 04 16 1 DAVbur 1 of thing.
2 The next step, beginning next month, is to 3 interact with what we see as the major issues, decay heat 4 removal and containment, those types of things, s 5 I have got a column here for the NRC expected 6 action date, where we give some feedback to DOE and its
.i 7 contractors.
8 And here in the last column on the right are the 9 interactions we had planned with the ACRS subcommittee --
10 today's meeting covering these. We have another session on 11 the HTGR scheduled in September to talk about the 12 containment confinement and what we will call the balance of 13 plant classification. That is the safety grade, nonsafety O 14 grade.
15 DR. SIESS: Those are the two items I mentioned, 16 but let me comment.
17 You have an ACRS briefing, and you say the 18 interaction with the ACRS subcommittee. That is a little j 19 different than what I said. I don't think interaction with l
- 20 the subcommittee is going to get you anywhere except in l
l 21 defining issues.
l l
22 I think there should be an interaction with ACRS, i
i i 23 and I think what you probably want -- and if you don't want l
24 it, you probably need it -- is an ACRS letter on the
_ 25 conceptual design.
1 I
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6850 02 05 17 1 DAVbur 1 MR. KING: Yes, we had planned for that.
I 2 DR. SIESS: That is not something that is going 3 to be accomplished simply by meeting with the subcommittee.
4 I would guess it is not going to be accomplished with one 4
5 meeting with the ACRS.
6 And so --
l 7 MR. KING: Let's talk about that a minute.
t 8 DR. SIESS: That schedule has got to be looked 9 at.
10 (Slide.)
11 MR. KING: This is the last part of that same 12 ' table. This covers FY '87 primarily, where we received the 13 PRA in September, we received the preliminary safety O 14 information document, and the thrust in '87 is to review the 15 PRA and the PSID. And out for ACRS we have TBDs.
. 16 One of the things we wanted to talk about 17 today -- two things. One is are there any additional .
18 interactions or briefings you would like to have in FY '86 19 on any particular aspect of design?
I 20 The second is when should we get the full 21 committee involved?
22 My own thought was we would get them involved 23 when we get the PSID, which will describe the design, and i 24 then have one or two or whatever number of sessions it takes 25 with the full committee because we would like a letter from 1
I J
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I 6850 02 06 18 j 1 DAVbur 1 them before we issue our SER on the PSID, which will be 4
2 later on in FY '87.
3 I agree with you, but we haven't worked out the 4 timing and the number of meetings yet.
5 DR. SIESS: I am trying to think to what extent 6 ACRS questions or comments might affect what is done in 7 completing the development of the reference concept, in 8 doing the PRA or the PSID. I am not sure you want to wait 9 until you have got the PSID and come in and get a lot of 10 questions that could have been addressed earlier.
11 It seems to me you want to get some kind of a 12 round with the full committee that has been brought up to 13 some level of understanding and see what kind of concerns 14 they have, then go back and see whether those concerns are i
15 being addressed or could be addressed in the PRA and the 16 PSID. That would expedite things.
17 That isn't as simple as it sounds because 18 sometimes it takes a document to get the ACRS attention. By
! 19 that, I mean both a document from the applicant and from the 20 staff, i
21 MR. KING: I guess what I would like to suggest 22 is maybe we could come back to this subject after you have 23 heard the presentations on the design and the licensing
- 24 approach and criteria. Maybe at the end of the day we could 25 talk about what additional interactions there should be.
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I 1 6850 02 07 19 1 DAVbur 1 I would appreciate your thoughts on what would be
(~x v 2 useful to the subcommittee and the full committee.
3 DR. SIESS: I agree.
4 (Slide.)
5 MR. KING: The last viewgraph is a quick summary 6 of the review support the staff is using to conduct these -
7 reviews over the next two years.
8 Right now we have a small contract with MIT.
9 That will be the fuel design and performance of the HTGR 10 fuel. We have a contract with Oak Ridge and Brookhaven to 11 perform independent analysis and assist us in the review of 12 the submittals from DOE.
13 Research, although they don' t have any money in O 14 FY '86, they have a lot of money carrying over from FY '85, 15 and they are completing work on an HTGR handbook, which is 16 primarily to document the state of the art in the HTGR 17 materials, fuels, designs, foreign designs, just trying to ,
18 get a good summary of the HTGR state of the art.
19 And then we are trying to maintain cognizance of 20 the foreign activities. Oak Ridge has contacts with the 21 German people involved in AVR and THTR, international 22 conferences -- Pete Williams went to Germany in September 23 for the International HTGR Conference -- cognizance of what 24 is happening with the THTR startup and the German HTR-500 25 design, the large gas reactor activities going on at this ACE-FEDERAL REPORTERS, INC.
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6850 02 08 20 :
1 DAVbur 1 point. r f) v 2 DR. CARBON: Which country seems serious on HTGR 3 activities at the present -- Germany?
l j 4 MR. KING: The Germans.
I
! ~5 DR. CARBON: Anyone else?
6 MR. KING: Japan is working on an HTGR, but its 7 timeframe I believe is the late 1990s. It is a small one.
8 I couldn't say more than that at this point.
t 9 With that I am going to turn it over to DOE, and 10 they will give you more details on the schedule and the 11 criteria for the design.
i 12 MR. MILLUNZI: The handouts that are coming
, 13 around, we have three-ring binders that you will be able to (1.) 14 put everyone's presentation in today so that when you leave 15 it will be in one package.
16 I would like to say good morning and to thank you 17 for this opportunity to meet with you today to brief you on l
l 18 the status of the DOE HTGR program.
I 19 It has been almost a year since we last met with 20 you on this subject. I think the last time was February l 21 Sth, 1985.
( 22 My name is Andrew C. Millunzi. I am the manager 23 of the safety and licensing for the advanced reactors at 24 DOE, both the HTGRs and the LMRs.
25 Today, of course, we are going to be talking on l
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J 6850 02 09 21 1,
1 DAVbur 1 the HTGR.
2 Before I get started, I would like to comment, we 4
3 were very interested in hearing your comments that you were 4 exchanging at the beginning of this meeting. I think you 5 will find that we have addressed every one of the items that 6 you mentioned, and we will continue to address them,
- 7 starting with the number of interactions.
8 I think when you look at our detailed schedule, 9 we fully agree that it is very appropriate and helpful for 10 us to have good contact with the NRC so that people can 11 understand the design that we are asking to get a license 12 for.
13 What you will find as we approach that, we want O 14 to make sure that you understand the design so that you can 15 pass judgment on the licensibility. However, we will talk 16 later about how we would like to have that interaction j 17 proceed.
l 18 One of our concerns is the roles and 19 responsibilities of an authority between both the applicant i
20 and the regulation, and we would like to talk about that a 21 little bit later.
l 22 Relative to the need of a containment and 23 separation of items such as Dr. Siess mentioned, we agreed 24 with that. Our approach, as you will see when we come to 25 it, we do not believe in attacking things by labels. We ACE-FEDERAL REPORTERS, INC.
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s
) 6850 02 10 22 1 DAVbur 1 are driven by requirements, and I think what we see is what 2 is the function that containment would have to perform? How 3 good does it have to get done, and how many ways could you 4 achieve that?
5 I think you will see that we are attacking that 6 problem because it is as important as Dr. Siess pointed 7 out.
< 8 As far as interacting with the ACRS, we will be 9 most anxious to have the interactions. Our TBD up on the 10 schedule that Tom King put up, we were hoping we could put 1
11 some dates on it as well as he did. So we would be anxious 12 to enter into that discussion with you at whatever time is 13 appropriate.
( 14 First of all, getting back to our presentation, I 15 really appreciate your agreement on the agenda. We 16 recognize that most people are very anxious to get at i
i 17 understanding the design.
18 In reviewing some of the past problems, we have 19 come to the conclusion that people have jumped to that step 20 too soon.
21 By that I mean we believe that it is vital to the 22 review of a design that this review be completed with an i
23 understanding from the top down of what the approach to 24 design and licensing was, what the requirements that the end 25 product is expected to meet, and how these requirements are J
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6850 02 11 23 1 DAVbur 1 to be met by the design, construction, and operation of the
\-) 2 plant.
3 So we think it is absolutely vital that you 4 understand these requirements, so that when we get the 5 design you can understand how that design evolved.
6 A key point that we will be making over and over 7 again, what we all have to do is assure that we meet 8 requirements. "
9 There are many ways to meet the requirements. We 10 need to draw upon the creativity of the design 11 organizations, the operating and the constructing operations 12 to be able to do that in the most efficient manner.
,_ 13 Therefore, what has to happen is they must be
\_/
- 14. given criteria. Then other people; namely, NRC, must be 15 evaluating how well those criteria are met by the way the 16 applicant is proposing.
17 The next job -- and it should not be in the mode 18 of telling them how to do it better -- next of all, then, it 19 is necessary to monitor them to make sure they are doing 20 everything they said the way they said they were going to.
21 So with that, our approach to this first briefing 22 with you and for subsequent ones will be from the top down.
23 24 25 O
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6850 03 01 24 1 DAVbw 1 (Slide.)
A) k- 2 We will always be relating things back to the 3 requirements and what it is one is supposed to accomplish.
4 For this meeting, the objectives of this briefing with you 5 are to brief you on our approach to the design. We would 6 like to brief you on the licensing approach and methodology 7 which has been proposed. We would then like to brief you on 8 the ACRS and brief you on the design status, and we would 9 like to receive your cour.onts, of course.
I 10 We're most interested in your comments on our 11 proposed licensing approach and methodology.
12 A fourth bullet that I've inadvertently left off
> 13 there, but it's most important, and I would like to read it
(:) 14 to you. The fourth bullet is, one of our objectives is to l 15 demonstrate that this concept is being developed in a ,
16 disciplined, comprehensive, cohesive manner by a 17 well-integrated capable team. We know that we can't 18 expect you to give approval of the safety that you hear, but 19 we fully expect that when we're through, that you will begin i
20 to appreciate the capabilities of this concept meet the l
l 21 safety and licensing requirements.
( 22 (Slide.)
23 I'd like to go back and review our program 24 objectives at DOE, and there are these which we have 25 presented to you in the paste.
1 ACE-FEDERAL REPORTERS, INC. l 202-347-3700 Nationwide Coverage 800-336-6646
6850 03 02 25 1 DAVbw 1 (Slide.)
2 The highly disciplined, capable, cohesive team 3 that I talked to you about is composed of these 4 organizations. What we have here is a unique combination 5 from the very outset, with a lot of interaction of the end 6 user, the developers and the vendors, as you can see from 7 this list 8 DR. SIESS: Where are the end users on there?
9 MR. MILLUNZI: The Gas-Cooled Reactor '
10 Associates.
11 DR. SIESS: Does that constitute your 12 constituncy of possible users?
13 MR. MILLUNZI: That is as of the present time, 14 Dr. Siess. Yes. This is the most visible one who is 15 providing support to the concepts. About 30 percent of tho 16 user utilities represent about 30 percent of the generating 17 capacity of the country.
18 DR. SIESS: Am I correct that GCRA has not
, 19 dropped its interest in the other forms of HTGRs and is 20 concentrating on the modular?
21 MR. MILLUNZI: I can't speak for HTGRs, to the 22 extent of having dropping. I think I can comment on the 23 word " concentrate," and they certainly have concentrated.
4 24 That is by far their main focal point. I think they would 25 always continue to look, but as of right now, for our ACE-FEDERAL REPORTERS, INC.
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6850 03 03 26 1 DAVbw 1 program with the GCRA Management Committee agreement, this i
2 is the primary focus to make this reference concept I 3 applicable. This team then, as was discussed with you last 4 year, completed a comprehensive, disciplined evaluation I 5 program that resulted in the selection of a reference 6 concept on which we will focus all our efforts.
7 We started out with over 16 concepts and came 8 down through this process to one, and that reference concept
- 9 is a 4 x 350 megawatt thermal HTGR plant that has an annular I
10 core of prismatic fuel in a steel vessel.
i 11 (Slide.)
4 12 A picture of this reference concept is here.
! 13 T will not spend any more time on this, because
' O 14 we will be presenting this in overview fashion very shortly 15 and in great detail during the closed session this 16 afternoon.
j 17 (Slide.)
l- 18 Next I would like to show you the overall 19 schedule of our activities to augment the schedule that Tom l 20 King gave you. This schedule is the schedule that is l
21 contained in the approved licensing plan that we have l 22 submitted and received and approval received from NRC.
I
( 23 That licensing plan is broken into these four 24 areas: a procedural approach, a technical approach, a j 25 design technology f amiliarization and. design technology '
($)
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j 2 We have submitted the leiensing plan, and it has I 3 been approved, and we're working with the Commission Staff 4 on this.
5 I'd like to comment, it has been a very, very 6 fruitful interaction that's going on from our standpoint and l 7 myself personally, I really commend Mr. Dircks for 3 establishing that Advanced Reactor Group. They have been 9 most useful, and I think we're really looking forward to
! 10 working with it. I think it's really going to help the i 11 process shorten up the time that we can get a license for y
4 12 this.
13 On the technical approach, we have submitted our i
14 proposed top level criteria, and we're anticipating getting 15 NRC response next month. We would be reaching, we expect, 16 towards the middle of the year, agreement on the licensing 17 bases. We have sufficient resources right now in the 18 budget, and we have taken into account Gramm-Rudman effects 19 in March. We will submit the PSID in the content and
, 20 quality that we had expected to at the end of September. We
! 21 also anticipate carrying out activities to be able to 22 complete these two milestones and get the licensability
- 23 statement in the end of fiscal '87.
t 24 So even more so, I think, not only is it that the l 25 Advanced Reactors Group at NRC is doing a very commendable i
l I
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6850 03 05 28 1 DAVbw 1 job in being thorough, but I think the whole interaction i
2 would suffer if that didn't continue.
, 3 (Slide.)
4 Here I'm going to have a series now to show you l 5 the extent of the interactions with NRC. Here these are. I
! 6 will not read them. They're all in the handouts. But you 7 can see here on the procedural approach we're starting with i
- 8 getting agreement on what the licensing plan will be and 9 taking into account the policy kinds of questions.
10 (Slide.)
11 On the technical approach interaction, we have 12 submitted the top level criteria. We will be talking to you 13 more about what we mean by this term " bridging." We have O 14 completed a method on selecting accidents, and I think this 15 will go a long way today to beginning to answer some of the 16 questions that Dr. Siess had brought up.
4 17 We agree that you need to do this, but we think 18 before you get engaged in that, you have to have an l
19 agreed-to selection method, and that's why we want to talk i
20 to you about that, before we get in too far.
21 Also we think it's important to look at this 22 question of safety-class selection. I think one of the 23 problems that the industry probably has is the use of that 24 term. We will be talking more about that as time goes on.
25 Also we are developing principal design ACE-FEDERAL REPORTERS, INC.
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6850 03 06 29 1 DAVbw 1 criteria. The principal design criteria are not.the same, 2 but they're akin to what people.have done in the lightwater 3 reactor area on the general design criteria, and we intend 4 to get back to the original intent of the general design 5 criteria, not to be confused with the general design 6 criteria, because they contain such a high level of 7 prescriptiveness and really aren't criteria. We have 8 changed the name, not to get confused.
9 And of course, we're here today with the briefing 10 to you.
11 (Slide.)
12 On the design technical issues, here is the item 13 here. There is just a brief discrepancy between this !
O 14 schedule and the one that Tom put it that we have 15 interchanged the decay heat removal and the fuel meetings.
16 But here is a list. Our whole program. And the team that 17 is back here. I have representatives from every team member 18 here. All our plans, detailed down to the month and day are 19 to provide these documents and meet these commitments.
} 20 We've even identified the dates inside these months that i
21 they will be met.
22 The point I would like to make here is that the 23 use of the word " issue" and how we will be addressing' them.
24 These items here all relate to functions that the plant has 25 to perform in order to meet the requirements. We will not
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'l DAVbw 1 be talking about these, as we go on, in what I might call 2 the classical sense of talking about any one of these things 3 in isolation, because it's improper, in our minds to treat 4 them in isolation. They are an integral part of an 5 integrated plant, but we will be going and addressing all 6 these items.
4 7 (Slide.)
8 Next on the design and technology 9 familiarization, we will be submitting a technology plan. I 10 look forward to your being impressed with the manner in 11 which that technology plan is being developed. We will be 12 performing the PRA. I might add here that FRA will be used 13 in this program extensively; however, we are going to employ 14 the broad definition of PRA. It is unfortunate that PRA, in l
15 many people's minds, has been limited to just being a risk 16 assessment item. Reliability engineering is what we will be 17 utilizing in our activities to a great extent, and a PRA is 18 a subset of using reliability engineering to assure yourself 19 that you have a quality product to meet the requirements.
20 The PSID, we have submitted the outline and 21 agreed to it with the Staff. The full submittal will be f 22 here. We will have a review of the preliminary safety 23 evaluation report next June and the licensability statement, 24 we're expecting next year.
25 The comments that were made about interacting l
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6850 03 08 31 1 DAVbw 1 with the Staff of ACRS, we will be raost anxious and very 2 willing to work out an aggressive schedule of those 3 interactions to assure that all of us can meet our 4 respective responsibilities by this date.
5 That concludes the introduction on the part of my 6 presentation.
7 DR. SIESS: fir . Millunzi, you've emphasized the 8 integrated approach, and you've emphasized a number of 9 differences from current practice. You mentioned the 10 principal design critocon versus the general design 11 criteria, the different approach to the PRA. I suspect
]
12 these are going to be areas where it's going to take some 13 effort to get the ACRS full Committee familiar enough with O 14 the differences to be able to follow this thing. This thing 15 will not be coming in in the conventional framework. We're 16 going to have to educate people to the differences, and
- 17 that's something that's just going to take a little more 18 time. I just mention that.
19 MR. MILLUNZI: Dr. Siess, I think the English 20 language, or my command of the English language is limited, 21 and the choice of words, I couldn't find one which was less 1
22 ambiguous. What I think you'll find is that we perform our ;
l 23 work in a very disciplined fashion and try to make sure that J
24 people are articulating exactly what they're talking about.
25 I do not believe that we will be doing things differently l ACE-FEDERAL REPORTERS, INC.
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-6850 03 09 32 1 DAVbw 1 to the degree that it sounds. What we will be doing
('T x/ 2 differently is that we will be clearly understanding between 3 all of us what is the standard for acceptability and how did 4 someone get there. That will be very clear, and it will be 5 very easy for people to find. We will know what has to be 6 done and how it has to be done.
7 DR. SIESS: But to get a 14- or 15-man ACRS to 8 have an equalli clear picture is not something you can do in 9 an hour.
10 MR. MILLUNZI: Right.
11 DR. SIESS: The Subcommittee may be able to help 12 you because of our familiarity with the other people on the 13 committee and their interests and concerns, but eventually, O 14 it's going to be 14 people that have to be brought up to 4
15 speed. I'm not pessimistic, just ralistic.
16 DR. CARBON: Another question. Sometime today,
, 17 someone will be covering the difference in detail between a 18 top-level criterion and a principal design one, and we'll be i 19 discussing what the technology plan is; is that correct?
20 MR. MILLUNZI: No. Later on today, we will not.
l 21 We are scheduled in there, as you can see, to present the 22 principal design criteria and submit them to the
! 23 Commission. We're not prepared to talk about those today.
24 We'll be more than happy to discuss them.
- 25 DR. CARBON: It wasn't so much to talk about them i
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4 6850 03 10 33 1 DAVbw 1 in detail, but rather to explain a little more what the
- 2 difference is.
3 MR. MILLUNZI: I think what the differences are 4 --
5 DR. CARBON: Will that be covered today?
6 MR. MILLUNZI: I can cover that right now and 7 tell you, in general, what they are.
8 The principal design criteria in the LWRs, in 9 many aspects, are very prescriptive. They're not criteria.
E 10 For some reason, the general design criteria have drifted 11 away from their original intent, and what we have tried to 12 do is go back to be consistent with the original intent of 13 the general design criteria, which is, you provide general 14 design criteria for a designer to meet, but you do not get 15 to be specific in there as to actually defining that 16 criteria.
s 17 DR. SIESS: At what point would you do that -- be 18 specific?
19 MR. MILLUNZI: At what time will be?
20 DR. SIESS: How many principal design criteria 21 can you have? Less than 58?
22 MR. MILLUNZI: The reason we're not prepared to 23 do it, we're in the process of completing it.
24 DR. SIESS: And you're working with the Staff?
25 MR. MILLUNZI: Right.
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- 6850 03 11 34 1 DAVbw 1 DR. CARBON
- But to a first approximation, the 2 principal design criteria compare with the general design 3 criteria.
4 MR. MILLUNZI: Right.
4
- 5 DR. C'.RBON : And the top-level criteria, what 6 would the be, briefly?
, 7 MR. MILLUNZI: If you would permit me, I'll come l 8 to that in a little while. And we have a presentation on 9 the top level criteria. Archie Kelly will handle that.
10 DR. CARBON: And the definition of the technology 11 plan?
! 12 MR. MILLUNZI: I don't understand the question on 13 the definition of the technology plan.
( 14 DR. CARBON: What is the technology plan meant to l
15 be?
i 16 MR. MILLUNZI: That is the plan of technology
', 17 efforts that will have to be performed to supplement the 18 technology that's not currently available to substantiate f
19 the conclusions that we reach in the license.
?
20 l
i 21
- 22 i 23 l
24 5
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6850 04 01 35 1 DAVbur 1 DR. SIESS: Mr. Millunzi, another question, I 2 guess on semantics.
3 I have heard the word " discipline," although it 4 didn't say a " disciplined engineering approach," which I 5 have heard frequently if not recently.
6 What I didn't hear were the words " defense in i
7 depth."
8 Is that part of your philosophy?
4 9 MR. MILLUNZI: I will cover that in my next 10 presentation. I am talking about discipline, and I seem to
- 11 have misplaced a viewgraph here. Ah, I have it. I will 12 talk to that.
J 13 We believe in defense in depth. I think the next I
i
( 14 part of the agenda was to get to our design and licensing 15 approach.
16 (Slide.)
17 This viewgraph depicts very completely our design
! 18 and licensing approach.
4 j 19 The blue part of this represents our approach to 20 designing a safe, reliable, economical HTGR.
I
! 21 The yellow represents the methodology by which we 22 will use to develop licensing bases unique for the HTGR in a 23 format that the Commission is familiar with dealing with.
! 24 So again we will be approaching and defining our i
25 end product, which will safe, economic, and reliable, and l ()
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! 6850 04 02 36 i
1 DAVbur 1 develop a methodology which will develop the licensing basis i
f' 2 unique for the HTGn in a form that the Commission is 3 presently accustomed to working with.
4 4 This is in recognition to the question you weee
, 5 raising, Dr. Siess, about how much reeducating, et cetera, 6 or changing.
7 Ideally, we would hope that eventually one of the ;
9 8 benefits from an advanced reactor program is that this kind 9 of an approach, which in reality is what all of us are 10 doing, would be adopted, but we are dealing with the world 11 as it is and not what we wish it to be.
! 12 Therefore, recognizing that problem that you are l 13 having, it is important for us, incumbent upon us to develop
( 14 this methodology, to make this bridge. Our presentation i
i 15 today will follow this.
16 Archie Kelley will present from the top down --
17 we are working this for you. He will present the' top level i
18 regulatory criteria proposed and the user requirements that 19 we use with the integrated approach.
20 When we are through, you will find that the i
21 integrated approach is nothing magic. In fact, if we are 22 successful in our interactions with you, when you are done 23 you will say he didn't tell me anything new. It is not 24 new.
, 25 All the integrated approach is, is a very 4
.(:)
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6850 04 03 37 1 DAVbur 1 disciplined way of describing how each of us does our job Q
k/ that we feel we are doing today. That is why the plants out 2
3 there today are safe.
4 Our plant here is going to be equivalently safe.
5 All of our comments and our need for developing this plant 6 have nothing to do with the level of safety in the present 7 operating plants. We believe they are safe for the American 8 public to use.
9 Then Tony Neylan and Bill Sherden -- Neylan is 10 from GA, and Bill Sherden is from Stone & Webster -- will 11 describe how the NSSS vendors and the architect engineers 12 take these top level requirements using this framework,
,_s 13 developing the engineering product.
U 14 Fred Silady then will say -- with this product, 15 he will describe the methodology that we intend to make this 16 bridge to be able to develop these unique licensing bases 17 for HTGRs in a format that NRC is familiar with.
{
18 I am a firm believer of defense in depth. Maybe 19 it is because all these years I keep telling somebody I am 20 19 years old and all of a sudden one of these days I am 21 going to realize I am really not.
22 That is why these plants are safe. They have 23 been built in a disciplined way and with defense in depth.
24 (Slide.)
25 That is why nuclear power -- I am a firm O
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6850 04 04 38 1 DAVbur 1 believer in it. So we believe in defense in depth.
2 We believe we are following it in a very, very 3 good way, which is going to provide excellent defense in 4 depth.
5 The way we do it is in pursuit of four goals:
i 6 First, our first goal is to maintain that plant 7 in safe operation.
8 Next, br-cause we are dealing with machinery and 9 with people, we are providing defense, that in the low 10 probability that you aren't able to maintain this safe
- 11 operation, we are able to protect the plant.
4 12 We need to protect that plant because that is the 13 investment part, and we have laid, as you will see, a very O 14 high requirement on how good we should do this job.
15 Then, consistent with nuclear energy, being the 1
16 only one who moves out and looks at the consequences of l
17 extremely low probability events happening in their 18 industry, wo do maintain control of the release of 19 radionuclides, and we are going to do that, as you will see, 20 in a very unique way.
21 Next, another aspect of nuclear power which is l'
22 unique to any industry in this country, we maintain 23 emergency preparedness in the event that in the extremely 24 low probability that we are unsuccessful at this level of f
25 defense.
l 1
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6850 04 05 39 4
1 DAVbur 1 Now, goal one is to be achieved by the highly ]
2 reliable operation of the plant with well-trained i
j 3 personnel. This will be made possible by providing a i
4 well-conceived design that is constructed to high quality 5 control and assurance standards. That is number one. That 6 is how we intend to get here.
7 We intend to use those high quality standards of 8 quality control and assurance in goal two and in goal three 1
i 9 to augment, that we will provide defense by utilizing the 10 inherent characteristics of the items in that plant and by 11 utilizing passive safety features.
.12 We intend to do these jobs so well that we will 13 place minimal reliance on four. We will have an emergency O 14 preparedness plan, but that plan will not call for 15 sheltering or evacuating from the general public after 425 16 meters from the plant.
i.
17 Therefore, we have imposed upon ourselves that we
! 18 maintain control of the release of radionuclides to the i
19 extent that any neighbor living more than 425. meters from 1
20 that plant his daily activities would not be affected and he
! 21 would not have to be evacuated or sheltered.
l 22 We believe all the things that are necessary to 23 achieve these are going to give us another type of safe L
24 nuclear power plant. In my mind, it is another type of l 25 nuclear power plant.
! ()
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a 6850 04 06 40 1 DAVbur 1 The safety of nuclear power plants isn't limited 2 by the coolant. It is how well you do these jobs.
3 DR. SIESS: Am I correct that in that last 4 objective you intend not only that the person near the plant 5 be protected from harm but that also he be protected from 6 f rig ht?
7 MR. MILLUNZI: Right. And also, going along with 8 that, I almost fell into the trap that I accuse other people 9 of, of being too complacent about another important item.
10 We from the very beginning are stressing 11 operations and maintenance, which result in putting in 12 safety in the beginning, and we are also making sure that we 13 do a superb job on protecting the personnel who are working O 14 in the plant.
15 DR. MARK: I really like the kind of things you 16 are saying. I am a little uneasy perhaps of the enormous 17 amount of having things come out the way you describe them.
18 It depends upon the management of the senior personnel of 19 the organization that is involved.
20 In what way and to what degree and by what means 21 do you think you can ensure that the people handling things 22 are handling them correctly?
23 MR. MILLUNZI: I think, as you will see, if I can 24 put this defense in depth back up here, we intend -- and as 25 I said earlier it is incumbent that when you go through ACE-FEDERAL REPORTERS, INC.
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6850 04 07 41 1 DAVbur 1 this that you have a full team from the very beginning and 2 that they be capable and that you have the end user, the guy 3 who is going to build and construct it and who is going to 4 operate and maintain and be involved in the program from the 5 very beginning.
6 We have gone to -- we think we have on this team 7 very, very capable people and capable organizations. We 8 will, as you will see throughout the development of this 9 program -- as you recall, when we submitted the response to 10 Public Law 96-567, which was weighed to recommendations on 11 how to improve the safety of LWRs, which I was responsible 12 for, our number one requirement there was related to 13 personnel, and we will be paying extreme attention to the O 14 training of everybody involved from the management down and i
15 their qualifications.
16 The plant is not going to have a safety problem 17 until it goes into operation and you have machines and 18 people, and it is necessarf, for example, that the 13 personnel, the maintenance and the operating people, all 20 part of Big O as we call it, Big Operations, are 21 well-trained and qualified.
22 So we will be paying close attention to that. We 23 will be trying to select, recruit, and keep people of high 24 quality.
25 That is why I used those words in my introduction ACE-FEDERAL REPORTERS, INC.
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6850 04 08 42
$ 1 DAVbur 1 so carefully. We really are dedicated to that. People are
< j
{}/ 2 going to make this thing work. )
3 DR. SIESS: Will this design have the same slow 4 response characteristics that they have, for example, at 5 Fort St. Vrain?
6 MR. MILLUNZI: Yes, and that, Dr. Siess, is right 7 here. We are going to use the inherent characteristics of
, 8 graphite and of fuel along with passive features. We are 9 looking for that all the time.
10 But that is right, we want to build on those kind 11 of characteristics. What we are trying to do is provide 12 time -- you know, one of the strategies here is to provide 13 time for the operations, the Big o, to make a decision and O 14 take action.
15 Now, you can't stop with that statement right 16 there. You have to define what actions and what decisions 17 they have to make. And as you will see, as we develop that i
18 we are concentrating on that.
i 19 DR. SIESS: Stepping ahead about three steps, l 20 when the license application comes in for one of these, can l
21 we assume that a training simulator will be an essential 22 part of it?
23 MR. MILLUNZI: Most of the guys on the team will f 24 die laughing right now when you ask me that question. I 25 have a number of red flag words, and one of them is ACE-FEDERAL REPORTERS, INC.
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6850~04 09 43 1 DAVbur 1 " simulator." I don't know what the hell anybody means when 2 they say " training simulator."
3 I will tell you what we will do. We will utilize 4 the latest technology in human engineering -- which is 5 another term I don't like but it is one that people are 6 familiar with -- we will use the latest techniques in 7 assuring ourselves that we get training. If it means using 8 some computer -- whatever it takes -- we will.
9 The word " simulator" bothers me because it means 10 different things to different people. But I think we do 11 need to have a device which is good for training people to
{ 12 identify and see how they would respond to things, help them 13 learn how they should respond to events, et cetera.
^
O 14 DR. SIESS: I don't understand why it means
} 15 different things to different people. I thought NRCL;had a
- 16 reg guide out that pretty well defined what a simulator 17 was and what it had to be able to do.
18 MR. MILLUNZI: I know, and I hope that as we go 19 on we can improve on that definition.
20 DR. CARBON: Andy, you are talking about high 21 quality people, highly trained, and so on, running the 22 plants, but you aren't going to have any say-so over that.
! 23 That will depend on the company out there'that owns and l
24 operates it.
l 25 It seems to me you almost ought to be aiming ACE-FEDERAL REPORTERS, INC.
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6850 04 10 44 2 o he e pe a ing 3 MR. MILLUNZI: Right. Well, one of the features 4 that you will see, an objective of the program, is to have a 5 plant which is fully automated. You will see that our 6 staffing goals are to reduce the staff.
7 We would like to get to the point -- I am in 8 agreement with you -- and our intent as a start to that is 9 to have a fully automated plant and to use, even with those 10 people, though -- I think it is incumbent for us to use 11 things -- like I am in complete agreement with Chet on using 12 things like the simulator, I think, as he has it in his 13 mind that it should be, to train them. You still have to I
O 14 train the people.
s 15 So an important part of our program is that we i
- 16 have the utilities involved fram the beginning, and one of 17 the tenets that we are trying to develop in this whole 18 program and the attitude that we are trying to develop in 19 this whole program is remember you are going to have to 2
20 train those people.
21 I am probably -- besides Dennis Wilkinson --
22 probably the biggest booster for INPO. We stress the 23 importance of INPO, the kinds of things that INPO is trying 24 to achieve, and the DOE program will try to lay down the 25 framework and the emphasis, and hopefully through that, in ACE-FEDERAL REPORTERS, INC.
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6850 04 11 45 1 'DAVbur 1 cooperation with the NRC, we can. develop that attitude.
ss ' 2 That is why it is important, when we get into 3 this, that you will hear us talking about we are beseeching
'4 you because we will be highly resistant to
, 5 prescriptiveness.
t 6 We need to get the people to understand that it 7 is their responsibility to design that plant right. It is 8 their responsibility to operate and to maintain it, and as 9 part of that fulfillment they must understand that plant.
10 They must be responsible for choosing the options and 11 getting them approved. They cannot rely on anyone else, the 12 NRC or DOE,-to tell them.
4 13 In the beginning, we should guide, but there is a
- 14 dif ference between guiding and telling. A problem that has 15 crept in, too, is that the guidance has become 16 prescription. We want to get back to the guidance that 17 relates to the criteria, for example.
18 So in answer to your question, we are striving l 19 mightily. The hidden agenda in this is to train the whole j 20 industry . hat they need that, and I woe,td like to add that 21 in my personal opinion it is not limited to the 22 owner-operators. The same thing applies to all of us.
- 23 So that is how we are attacking it, and I don' t i
- 24 know of any other way to do it. In the end they are 25 responsible. Therefore, we have to work together to make ACE-FEDERAL REPORTERS, INC.
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6850 04 12 46 1 DAVbur 1 sure they are capable of fulfilling that responsibility.
,s
' -)
4 2 I thinx they are responsible now, looking at the 3 great safety record for nuclear power plants. However, when 4 one looks at improving -- my iefinition and the reason I l
5 use that word is not that improvement means what is there l 6 isn't good enough. My thing is all of us can be improved, 7 and that is what we are looking for -- is for improvement.
8 And I am personally -- the number one item is 9 personnel.
10 11 12 13 O 14 15 16 17 18 19 20 21 22 23 24
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6850 05 01 47 1 DAVbw 1 DR. SIESS: Max, I'm not sure I agree with you 2 that I want a plant that any old guy can operate. I haue a I 3 feeling I'd be more worried about com,nlacency than I would 4 about incompetence.
I 5 DR. CARBON: My point was, DOE doesn't have any 6 control over the people.
7 DR. MARK: I'm still left a little bit, not i
8 really uneasy, but if there was such a thing as a sleazy 9 management, then all the things you're saying, don't
{ 10 necessary apply, and I don't see any way of assuring that we 11 don't run into a sleazy management.
12 MR. MILLUNZI: Along that line, Carson, we are l
, 13 trying to build a just reward system, okayt? And it goes
(:) 14 the following way:
a I 15 This plant, by what we're going to do in 3 and'4, 16 will be such that anybody 425 meters from that plant, will 17 not be injured. The guy is sleazy is going to lose his
[ 18 shirt, because his plant is down and out.
19 So that's our ace in the hole against that kind j 20 of thing.
I 21 DR. CARBON: I don't want to prolong this, but I J
22 thought, that's just contrary to what you said in Item 4
23 No. 2.
i 24 MR. MILLUNZI: No, Item No 2 here is here, if I 4
25 could, I'd like to defer that to we're all dotte.
! (:)
4
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6850 05 02 48 1 DAVbw 1 DR. CARBON: Maybe we'd better, because I think fs
- 2 we're falling pretty far behind schedule.
1 3 (Slide.)
4 MR. MILLUNZI: Lastly, in the summary, then, what ;
5 we're doing is, we are identifying the criteria, and that is 6 done. We're awaiting the final approval. We are developing 7 the process to derive the licensing bases unique for this, 8 which NRC is familiar with. We have briefed and met with 9 the Staff, and we will be submitting our documentation in 10 the next several weeks. And then we take this an apply the 11 process to identify these licensing bases, and these are in 12 process.
13 Archie Kelley will follow me. He will be 14 talking about these criteria. Then Tony Neylan and Bill 15 Sheridan and I will talk to you about this approach.
16 Neylan and Sheridan will talk about the product that results 17 from that, and then Fred Silady, with his presentation, will 18 cover points 2 and 3.
19 So with that, I'd like to turn it over to Archie 20 Kelley from the Gas-Cooled Reactor Associates.
21 MR. KELLEY: Good morning. My name is Archie 22 Kelley with Gas-Cooled Reactor Associates.
23 As Andy told you, Gas-Cooled Reactor Associates 24 represents roughly one-third of the nation's generating
- 25 capacity. Our members are vitally interested in the tj ACE-FEDERAL REPORTERS, INC.
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6850 05 03 49 1
1 DAVbw 1 development and support the development of the gas-cooled 2 technology in-this country. Also to confirm what Andy told 3 you, our surveys of our member utilities, as well as 4 potential process heat users, indicates that this concept-5 you're going to hear about today, indeed, is an excellent 6 match to our utilities projections of their future energy 7 needs in the next several decades.
I 8 In addition, we think it's an even better match i
9 than the former large plants we worked on for the potential I'
10 process heat and higher temperature markets which may be out j 11 there in the future for nuclear power applications.
! 12 To move into my topic today, as Andy told you, 13 we're going to use this as our road map for the presentation 4
} 14 today.
)
~
15 My subject matter is to brief yo on the top level j 16 user and top level regulatory criteria that we feel 17 appropriately fill.these top two boxes on the left-hand side i 18 of this diagram.
! 19 (Slide.)
20 I'm going to do this in three parts. First, very 21 briefly, the purpose of these requirements, some of the I
22 ground rules by which we develop them.
1
- 23 Secondly, I'll tell you what the utility user l 24 requirements are that are driving this design you're going 2
25 to see today, and then finally, most importantly, describe
! }
l
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4 6850 05 04 50 1 DAVbw 1 to you the regulatory criteria that we have proposed to the 2 NRC Staf f to use to evaluate our conceptual design. And in 3 doing that, we came up with some bases for selection which 4 I'll identify and then the proposed criteria themselves.
,! 5 (Slide.) ,
6 First of all, with regard to purpose. The 7 purpose of these top-level criteria and requirements, quite i 8 simply, is to help us define how safe and how economical we 1
9 want the power from this plant to be.
10 As Andy has told you, working from the top down, t 11 we have defined four goals that we will allocate these i
12 requirements and criteria to. Goal 1 being maintain safe 13 plant operation. Maintain plant protection is Goal 2.
i
! 14 Maintain control of radionuclide release, Goal 3. And 15 finally, maintaining emergency preparedness.
4 16 So our goals, you will see, are allocated against 17 -- excuse me. Our criteria are allocated against the 18 structure of four goals, which we believe are those goals .
19 which are necessary to maintain safe economical power.
l 20 (Slide.)
i
! 21 Moving then into the second topic, the utility l 22 user criteria. Here, of course, I will be emphasizing the i
23 aspects of economics, because that's primarily what the user j 24 is focused on. You will see, of course, that there was a i
- 25 crossover by the user. He is also interested in the safety l
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'6850 05 05 51 1 DAVbw 1 aspects of this plant, but my emphasis from the user 2 viewpoint'will primarily focus on the question of how 3 economic, beginning with some definitions from an overall 4 viewpoint of what economic means to our utilities. We have 5 some numerical goals that have been defined, first of all.
6 We have a target that we show a 10 percent economic 7 advantage over state-of-the-art coal plants. We consider 8 that the appropriate competition for this plant at this 9 time.
10 We defined a siting envelope we want this 11 economic criteria to be met, considering a siting envelope 12 which cover roughly 85 percent of U.S. sites and other 13 criteria, such as the service life of 40 years, and such O 14 things as would cover the overall plant.
j 15 Moving then into the criteria for maintaining l 16 safe plant operation, the overall criteria from the user i l 17 viewpoint is that we show an equivalent unavailability owing 18 to planned outages of less than 10 percent.
j 19 You will recognize that to meet this, there are a
- 20 number of suballocations that relate to such things as how i
3 21 we maintain and operate this plant, and there are safety
! 22 considerations with regard to the worker exposure, and so I
- 23 forth, in order that we might meet that target. And those 24 sublevel criteria are, in fact, identified by the user.
25 DR. SIESS: Excuse me. In that economic ACE-FEDERAL REPORTERS, INC.
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l 6850 05 06 52 1 DAVbw I advantage over coal, what does that assume about the back 2 end of the fuel cycle?
3 MR. KELLEY: This does not assume, obviously, the 4 recycle fuel cycle.
5 i DR. SIESS: No recycle?
6 MR. KELLEY: No recycle.
7 DR. SIESS: And permanent disposal of the 1 mil 8 per kilowatt hour?
9 MR. KELLEY: That's correct. We have a set of 10 user defined economic ground rules, and we're evaluating 11 that. And that is an example of one of the elements in the 12 ground rules.
13 DR. SIESS: I just wanted to make sure. It O 14 wasn't really pertinent to this.
! 15 (Slide.)
16 MR. KELLEY: In the area of the second goal, 17 which is maintain plant protection, there's a lot of user 18 interest, and therefore, we have a number of requirements in 19 this area that are imposed on the design. Some of the key 20 ones I've shown up here.
i 21 First of all, in terms of maintaining plant 22 protection, the user is concerned about those events that, 23 in terms of equipment damage, may be relatively trivial, but 24 may result in cumulative plant outages that impact to the 4 25 overall availability. We therefore have a goal that the
(
i i
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6850 05 07 53 1 DAVbw 1 equivalent unavailability, owing to such unplanned equipment 2 outages be less than 10 percent.
3 You can take this 10 percent and add it to the 10 4 percent planned outage criteria, which I showed you 5 previously, and we have an overall goal that the 6 availability be greater than 80 percent, the equivalent 7 availability.
8 Moving down, in terms of level of concern, for 9 more serious events, you are obviously concerned about i
, 10 equipment damage. We therefore had a criterion basically
{ 11 that the annual expected value of damage to equipment be 12 less than the insurance premium which we use in our cost 1
13 estimate of $4.5 million a year. Obviously, we're not O 14 talking about the expectation of an annual -- in fact, an 15 annual loss of $4.5 million, but overall events that might 16 impact equipment in the plant, we're asking that on the 17 average, they be shown to be less than S4.5 mill. ion.
18 Finally, moving down to even more serious 19 potential events in the plant, those, in particular , might 20 be of nuclear significance and might have public 21 repercussions. The user, of course, is concerned, 22 additionally, not just because it might impact his plant, 23 but because, in fact, it might impact the entire base of 24 installed HTGRs in a population of HTGRs in the future.
25 Therefore, there's obviously quite an aversion to such ACE-FEDERAL REPORTERS, INC.
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l l
6850 05 08 54 1 DAVbw 1 events. That's shown by our criteria that we require, that p
v 2 the designers show that there be a mean likelihood of a loss l
1-3 of a single reactor, be shown to be less than 10 to the
- 4 minus 5th per year.
( 5 DR. CARBON: This is total loss? Wipeout?
6 I mean financially and operability?
7 MR. KELLEY: This effectively would be a loss of 8 reactor so serious that it would be a write-off, as far as 9 the utility was concerned.
j 10 DR. SIESS: A write-off of one reactor out of j
! 11 four?
4 12 MR. KELLEY: That's right.
l 13 DR. SIESS: But not such that it would disable 14 the whole plant?
15 MR. KELLEY: The concern is obviously, if this l 16 were a nuclear-related event, again, that it could have f
17 public repercussions, that even though the other plants are j_ 18 not damaged, they could be commonly affected.
! 19 (Slide.)
20 Finally, as Andy has already told you, we do havs
{. 21 a key user requirement in Goal 3, relating to maintaining s
i 22 control of radionuclide release. The user is requiring that l 23 we do this so well with this plant, that we would, indeed, 24 meet all the top level regulatory criteria without taking
{ 25 l credit for sheltering or evacuation of the public, such that LO 1
L i ACE-FEDERAL REPORTERS, INC.
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l 1
4 6850 05 09 55 i r l 1 DAVbw 1 the utility user need not plan for the offsite evacuation j l -
2 and sheltering of the public. This currently has been 3 interpreted, and we believe quite conservatively in the 4 design process as meaning that we show that we meet the 5 protective action guide doses for sheltering. Mainly, 1 rem 6 whole body and 5 rem thyroid for events that have assessed i
7 mean frequences of greater than 5 x 10 to the minus 7 per 8 year.
9 DR. CARBON: What was that again? I rem whole 10 body and 5 rem --
i 11 MR. KELLEY: 1 rem whole body, 5 rem thyroid.
I 12 '
DR. SIESS: That last statement, is that i' 13 equivalent to saying that the probability of exceeding the
' C:) 14 PAG is 5 x 10 to the minus 7 per year for all events?
l 15 MR. KELLEY: It would be less than 5 x 10 to the I 16 minus 7 per year for all events.
- 17 DR. SIESS: Okay, for all events. Okay.
f 18 How did you arrive at the 5 x 10 to the minus 77 19 Working backwards from the safety goal?
20 MR. KELLEY: In effect, you'll see further 4 21 discussion of this in Fred Silady's paper, but in effect, we 22 did work back from the safety goals.
l 23 DR. SIESS: Is the 10 to the minus 5 then l 24 equivalent to a core melt criterion?
i 25 MR. KELLEY: Yes, but again, it's from the i ACE-FEDERAL REPORTERS, INC.
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i 6850 05 10 56
? l- DAVbw 1 viewpoint of equipment damage that would have a nuclear y
'] 2 consequence. Core melt doesn't have much of a meaning for a 3 graphite core.
) 4 DR. SIESS: I said equivalent to that.
4 5 MR. KELLEY: In loose terms; yes.
.i 6 DR. SIESS: So it would damage the plant 7 severely.
8 MR. KELLEY: And have offsite public 9 repercussions; yes.
10 DR. SIESS: But not offsite doses?
11 MR. KELLEY: But not significant offsite doses.
12 Again, what we would be concerned with is events that would 13 cause the public concern, even though they would not O 14 actually necessarily evolve health effects.
15 MR. MILLUNZI: I might add, Archie, in order to 16 help the situation, the design also has to be adverse to 17 long outages, six months.
18 MR. KELLEY: That's a point. You're not seeing 19 all the tiers of requirements that we have, but in terms of 20 the 10 percent unplanned outage, the user has an aversion to 21 very long outages, namely, those greater than six months.
22 They can only contribute 10 percent to that 10 percent.
23 DR. SIESS: That would be something like 24 Davis-Besse?
25 MR. KELLEY: Yes; that type of thing.
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6850 05 11 57 1 DAVbw 1 DR. SIESS: You would expect that something like 2 that would affect all four units; right?
3 It would be a plant shutdown, not a reactor 4 shutdown?
5 MR. KELLEY: That's the concern for the 10 to the 6 minus 5th criterion.
7 MR. MILLUNZI: That's a long outage. . It ' s pe r i
8 reactor.
9 DR. SIESS: Well, I would be willing to bet that i
10 if you had something that the NRC considered like 11 Davis-Besse, it would shut down the plant.
]
12 MR. MILLUNZI: That's right ; yes.
13 DR. SIESS: We're talking about a regular O 14 shutdown now.
, 15 (Slide.)
16 MR. KELLEY: Let me move into my last topic, the 17 most important one here, and discuss with you the approach i
18 that we proposed the NRC Staff with regard to top level 19 regulatory criteria for this concept. Obviously, it's not i
f 20 our role nor were we really interested in proposing unique
! 21 or new top level regula~ory criteria for this concept. But 22 we were concerned that in the current body of regulatory 23 criteria, Code of Federal Regulations, et cetera, that were 24 a number of criteria that we~e rather unique to lightwater l g- 25 reactors, somewhat restrictive in terms of lightwater i
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i l
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6850 05 12 58 1 DAVbw 1 reactor design features, a we felt it important to identify 2 those criteria that were clearly generic versus those that 3 appeared to have a lightwater reactor history to them or 4 lightwater design background to them.
5 To do this, we established some selection bases, 6 which I've shown on this Vugraph, and we hope that these 7 selection bases would help us in this weeding process to 8 identify criteria that, indeed, were truly generic.
9 The first basis for selection that we used were 10 that the criteria must be direct statements of acceptable 11 consequences or risks to the public or environment.
12 We believe clearly that requirement was 13 independent of the design.
14 Secondly, quite clearly, that the criteria should 15 be independent of the plant design. It was mentioning 16 specific plant design features, and so forth. Obviously, 17 that would not be the case.
18 Thirdly, in order to be useful to us in our 19 design process, we believe the criteria must be 20 quantifiable. Just as I've shown you, the user criteria 21 were quantified. We believe the regulatory criteria used in 22 this conceptual design process need to be quantifiaole, in 23 order to provide appropriate guidance to the designer.
24 So with these bases, what we did, we scre'.ied the 25 requirements and criteria and guidelines that we find in the ACE-FEDERAL REPORTERS, INC.
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6850 05 13 59 2 DAVbw 1 regulatory documents and then identified those that we
()
k
- 2 thought, in fact, were generic.
3 (Slide.)
4 Starting then again with the statement of overall 5 safe economic power, we felt, obviously, the best candidate 6 for this was the guideline criteria found in NUREG 0880, 7 namely, the interim safety go.'Is. Namely, there, the 8 criteria that met our selection bases were the individual 9 and societal mortality risks that are identified therein and 10 also the cost-benefit criteria, which are identified 11 therein, that would be invoked only if the mortality risk 12 criteria were not met.
- 13 Obviously, we intend to meet these criteria, and
'"' therefore, based on the interim safety goals as currently 14 15 stated, we would not need to invoke cost-benefit criteria.
16 17 18 l's 20 21 22 23 l 24 25 O)
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6850 06 01 60 1 DAVbut 1 DR. SIESS: Are you familiar with some of the 2 proposals for final safety goals?
3 MR. KELLEY: We have been following that, 4 obviously, with great interest. Once those are promulgated, 5 we will obviously have to consider those.
_6 I might note, one thing you will see j 7 conspicuously absent here in NUREG-0880 is -- of course, 8 there is an identification for a design performance goal for 4
9 large scale core melt. However, we felt that that did not i
i 10 in fact meet any of the three selection bases which I have I
11 identified to you previously. It is not a direct statement i
12 related to public health and safety.
13 We don't believe it really is independent of i 14 plant design. We believe it is light water reactor only, 4
15 and it is in terms of understanding the consequences of core l 16 melts and light water reactors. And finally, because of the j 17 terminology, it is not really something we can quantify on a
- 18 graphite reactor.
. 19 DR. SIESS: I agree with you 100 percent, but I l
20 think you have got some people to convince.
21 MR. KELLEY: Yes.
22 DR. SIESS: Not only have core melt criteria or
! 23 core damage criteria begun to assume a primary role, but 24 there is a move toward including descriptive containment 25 performance criteria, which is a particular problem.
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j 6850 06 02 61 i
! 1 DAVbur 1 MR. KELLEY: Yes.
7 2 Recognize, however, as I showed you previously, i
j 3 from a user perspective in any case we do in effect have a 4 performance criteria related to damage to one of the
! 5 reactors.
l l 6 So obviously, that is not neglected in our design l
7 process. We just feel it is coming appropriately from the
- 8 user.
9 DR. SIESS: But when you say a user requirement, 10 it means you hope that it wouldn't be an NRC requirement? .
- 11 - MR. KELLEY
- No. We would hope it would be a j 12 user requirement.
i 13 (Slide.)
14 Going through each of the sublevel goals in terms
. 15 of maintaining safe plant operation, we believe obviously F
l 16 the candidates there are the permissible dose levels and i
i 17 activity concentrations for unrestricted areas, which are 18 contained in 10 CPR 20.
j 19 Analogously, the numerical dose guidelines i
l 20 contained in 10 CFR 50, Appendix I. In goal two, obviously l
! 21 that would apply as well. But there are obviously concerns l 22 related to occupational exposure criteria. So those would l 23 be invoked under goal two.
l
! 24 (Slide.)
i 25 Under goal three, related to maintaining control
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i I 6850 06 03 62 j 1 DAVbur 1 of radionuclide release in the event of off-normal or 2 accident conditions in the plant, we are invoking 10 CFR, 3 Appendix I doses applied on an expected value basis to 4 events which might occur in the plant lifetime.
] 5 In other words, by taking the sum of the product 6 of the frequencies and the consequences, we show that we
- 7 meet 10 CFR 50, Appendix I, offsite dose values.
i 8 For accident events, obviously we have invoked j 9 the numerical dose guidelines contained in 10 CPR 100.
10 Finally, with regard to the goal of maintaining
! 11 emergency preparedness, we believe that the criteria should i
12 be the criteria contained in EPA-520, the PAG doses 13 contained therein, which are guides in terms of triggering 4 ) 14 offsite actions such as sheltering and evacuation.
j 15 This completes my summary of the top level i 16 regulatory criteria. ~
l 17 I recognize that we have submitted a document to
- i t
l 18 Tom King's group which identifies our proposed criteria in l 19 some detail. It also identifies the selection bases which I l
l 20 described to you as we understand it. This is currently I
j 21 being reviewed by the NRC, and we expect to hear from them i
j 22 shortly on that.
23 DR. SIESS: I have a question I forgot to ask.
24 It may have been answered.
25 On one of the slides you said the siting l ACE FEDERAL REPORTERS, INC.
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I 1
4 6850 06 04 63 1 DAVbur 1 envelope covering 85 percent of U.S. sites.
O 2 Can 1 assume that that is primarily seismic?
. 3 MR. KELLEY: It covers a number of things -- -
I seismic, wind loading, temperature. For example, Andy 4
5 mentioned to you we have a criterion of 425 meter exclusion l
l 6 area boundary. That number actually came by enveloping 85 4
.7 percent of the sites.
8 DR. SIESS: Existing sites?
9 MR. KELLEY: Existing sites in that case, that is 10 correct, existing or sites under construction. 425 meters 11 our assessment tells us bounds 85 percent of current sites.
12 DR. SIESS: But in terms of potential sites,'the
, 13 only thing to exclude would be certain seismic regions,
(:) 14 right?
15 MR. KELLEY: Yes.
l
- l. 16 DR. SIESS: If you chose a state, you couldn't 17 choose California?
j 18 MR. KELLEY: Certainly, there are parts of 19 California which would be difficult.
6 20 DR. SIESS: And a few in Wyoming.
i 21 MR. KELLEY: Okay.
, 22 (Slide.)
! 23 Putting our road map back up on, then, the next j
{ 24 topic.
l 25 Now that you know what the top level criteria and i
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6850 06 05 64 1 DAVbur 1 user requirements are that have been defined, Andy Millunzi i
{~T s/ 2 will describe to you the integrated approach technique we 3 use in a system engineering sense to meld these requirements 4 - into a design.
5 MR. MILLUNZI: I am going to cover, as Archie 6 pointed out, a description and a discussion on the i'
7 integrated approach.
8 -First of all, when one has that, if you are going
- j. 9 to set out to do a job like this you have to know what are 10 the purposes and benefits of doing this.
11 (Slide.)
12 The purpose of the integrated approach is used to 13 develop requirements. It is used to evaluate the options O 14 that one can develop to meet the options, and it is to 4
15 communicate with reviewers.
16 By this, I mean that the first point for a 17 nuclear power plant, requirements have to be developed which 18 will respond to the top level user and regulatory 19 requirements.
l '20 So taking those requirements with the integrated 21 approach, the purpose of it is to allocate, to develop and
- 22 allocate those requirements. Then it is to evaluate the i
s s 23 options that the design, construction, and operation 24 institutions propose to meet those requirements.
25 And then a purpose for it is to communicate with
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6850 06 06 65 1 DAVbur 1 all kinds of reviewers. It does not mean the reviewers at
([-
2 NRC. It is not restricted to anybody. It is to be able to 3, cJmmunicate what this plant is doing and why it has its 4 capabilities.
5 (Slide.)
6 The benefits from the integrated approach are 7 many.
8 In thinking about our presentation last night, I 9 have added a few viewgraphs which aren't in your package, 10 but I am sure we can get copies to you.
11 The benefits from it are that the development of 12 this integrated plant model provides insight into the design
,_ 13 structure of the plant in terms of definition of functions, .
U 14 definition of a minimal set of components which af fect each 15 function.
16 This will lead to a better understanding of the 17 plant in that the exact reasons for having each piece of 18 equipment are identified. That is a very important item, to 19 understand why every piece of equipment is in there.
20 Every piece of equipment is in there to perform a 21 function. This may assist in defining what is or is not 22 functionally required to be safety related. It will give a 23 definition of which components are most critical to plant 24 operations in terms of economics and safety. )
l 25 (Slide.)
(/ )
l 1
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6850 06 07 66 1 DAVbur 1 Continuing on, it provides insights into the ss 2 importance of human interactions, which are knowledge-based l
3 behavior, rule of skill-based behavior, or institutional or 1 4 administrative.
5 Three, it provides a framework for information 6 processing techniques which allow the human to perform a 7 top-down analysis of any plant condition, determine the 8 plant status as expeditiously as possible, and thus reduce 9 the time required to arrive at a decision and reducing the 10 likelihood of making a wrong decision.
11 Back to our emphasis on how important the people 12 are and their training.
13 (Slide.)
n' 14 Continuing on, it provides an analysis which will 15 indicate the importance of institutional contributors which 16 can then be compared to their role in the existing plant 17 function administration. This may lead to areas which may 18 require emphasis or impr oveme n t.
19 Continually, you need to be operating these 20 plant, to continually review yourself, learn from yourself, 21 and how do you identify where these improvements can come 22 from.
23 And five, it provides a functional model for an 24 optimal offsite emergency preparedness which our emergency 25 plan could be evaluated against.
O x_/
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6850 06 08 67 1 DAVbur 1 (Slide.)
(3 k/ 2 The benefits, continuing, the integrated approach 3 is expected to minimize the total cost of power by having 4 all of these attributes and having the interactions. We 5 would have a shorter licensing time, a shorter construction 6 time, be able to minimize retrofits and minimize operating 7 and maintenance costs through the better understanding of 8 things before and after the plant is put online.
9 (Slide.)
10 In the design areas, the savings are due to a 11 clear understanding by the designers, contractors, and 12 operators of what their roles and responsibilities are.
_s 13 The bringing online and the operating of a plant 14 requires many organizations and many subunits inside these 15 organizations, and you need to pay attention to interfaces 16 so that you don't have somebody building an engine for your 17 airplane that is too light for the payload that it is going 18 to take off. That is what happened to Westinghouse, for 19 example, in that area.
20 And early identification of interfaces which 21 reduces the risk of later more costly revisions. That is 22 part of the point I think, Dr. Carbon, that you were talking 23 to.
24 It gives a visibility of the basis for the design 25 requirements. This is very, very important. It would help 7_
i V I
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6850 06 09 68 1 DAvbur 1 eliminate unjustifiable retrofits. ,
2 Lastly, we expect to get savings due to how it i 3 will affect our development programs.
4 Now, I will get into -- we needed this kind of a 5 framework, we found. We needed something which is the
~
6 integrator. So I am going to give you a very brief 7 description of the integrated approach.
8 As Dr. Siess says, it will take some more 9 interactions for people to become more familiar with it.
10 However, you will find, as we talk to you today, and
! 11 hopefully and assuredly as we get to talk in more detail, 12 that we are not talking about anything very sophisticated.
13 We are not talking about something that people don't do O 14 already.
i 15 The difference is we do it in a very disciplined, 16 organized fashion. We articulate very clearly what it is we 17 are doing, and we provide the traceability and the i
18 visibility.
- 19 So we have put in systems engineering way how i
I 20 everybody believes they are doing their job. Now to do this 21 and to get all these benefits, we discovered you have to go 22 back to square one. All of us have to go back and say from 23 the top down, why are we here, what are we after?
, 24 (Slide.)
t 25 Our objective, all of us, is to use the heat from ACE-FEDERAL REPORTERS, INC.
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6850 06 10 69 1 DAVbur 1 a nuclear reaction to produce electricity safely, 2 economically, and reliably. Therefore, the owner of the 3 plant, if it wouldn't be for the radionuclides that his 4 product possesses, he would only have to be concerned with 5 meeting the economic criteria.
6 Because his product contains a toxic or 7 radioactive material, society has imposed a regulatory body 8 on him whose responsibility is to protect the health and 9 safety of the public.
10 So the owner-operator of this plant has to now 11 use a plant which he could sell at a profit electricity, but 12 he also has to meet regulatory requirements. l 13 Now, with that dual set of requirements he has to
. O 14 try to figure out how he is going to approach that. There 15 are many ways that those requirements can be met.
16 As I said earlier, he has to be afforded the
! 17 ability to develop the option as to how he wants to do it.
- la The people who are lending him the money have to evaluate if i
l 19 it looks like it is economically viable, and the regulatory l
20 bodies have to see if it is adequately protecting the health 21 and public safety.
22 But both of those reviewing organizations, then, 23 have the responsibility to evaluate if his proposed way is 24 good enough. They must resist telling him how to do it.
1
- 25 The way in the nuclear power business we have l ()
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4 6850 06 11 70
'l DAVbur 1 elected to do it is --
2 (Slide.)
3 -- we first of all develop a design and we 4 construct and operate it so that it will stay in the normal 5 operating envelope.
6 (Slide.)
7 Recognizing that this plant has people and 8 equipment involved, we then provide features, both in 9 operation and training, et cetera, so that we will maintain 10 the plant protection.
11 (Slide.)
12 Then, knowing that we have this radionuclide 13 content in the plant and we have to meet regulatory O 14 requirements, we in our obligation to society have to be 15 able to -- in the event that the damage gets to be so severe 16 that we are releasing radionuclides, we have to maintain 17 control of those. radionuclides so that they do not injure 18 the health of the public.
l 19 (Slide.)
20 Then, last of all, the nuclear industry again, i
j , 21 unlike any others, has to provide for emergency 22 preparedness. It turns out that it is all of these 23 activities that the utility, the designers, and everyone 24 else has to perform adequately enough to meet these 25 requirements.
i l
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6850 06 12 71 1 DAVbur 1 There are many ways that this can be done, and so 2 it is the responsibility of the applicant and his suppliers
- 3 to develop a way to do that.
4 How we do that is nothing new. This is how all 5 -engineering jobs do it.
i 6 (Slide.)
i-7 What we have here is a goal. In order to achieve 8 that goal, there are functions that have to be performed.
e i 9 In this viewgraph I only have one, but there are a number of 10 functions which go together that have to be completed in 11 order to achieve the goal.
12 As engineers, what you do in a systematic way is
! 13 we take the function, and a function is modularized. You
(~)
%./ -
14 have subfunctions. For each of these subfunctions you have 15 systems and components which, together with human actions, i
16 enable you during operation and maintenance to achieve the 17 functional requirements of the subfunction which, taken 18 together, enable you to meet the functional requirements of 19 the function, and the functional requirements of all the i
20 functions accumulated enable you to meet the requirements of 21 the goal.
22 So as you can see when you develop this, you have j
23 what we call a tree. So what we then perform is a 24 functional analysis of this tree to try to determine, one, 25 what are the functions, what are the subfunctions that have
()
l l
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6850 06 13 72 l l
1 DAVbur. I to be performed, then define and allocate from the goals how
> ) 2 good that has to be done. That is what functional 3 requirements are. I l
- 4 The functional requirements at each level are )
I
- 5 then further allocated down to the subfunctions. It is i
6 against these requirements, then, using design criteria, 7 that the systems and components get designed along with the i 8 humans to meet all of these items.
9 It is important that in many aspects right now i
10 the visibility and the traceability -- there is improvement i
11 that can be done in that. So if from the very beginning we 12 are worrying about traceability, we have developed a 13 documentation logic which will preserve for us at all times 14 for everyone to see why did we do what we did.
15 16 e
- 17 18 '
i 19 20 21 22 23 24 4 25 1
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6850 07 01 73 1 DAVbur 1 (Slide.)
'k / 2 This is accomplished by having these sets of 3 documents.
4 Starting from the top, there is an overall plant 5 specification. In that document are user requirements, NRC 6 requirements.
7 Taking all of these, we perform plant level 8 analyses, trade studies, design studies, and out of this we 9 define the system requirements.
10 These system requirements then are passed down to 11 the system design description, the SDD. These become 12 requirements.
13 The system designer then has the opportunity and 7_
\'"] 14 the responsibility and the option to develop options as to 15 how those requirements will be met. They do the same thing 16 as what was done up at the top. They perform system or 17 subsystem level analysis, trade studies, and design 18 studies.
19 Out of that will fall out design selections.
20 These design selections in turn are then passed down as 21 requirements to the lower document.
22 For example, from the system, the system design 23 selections are then, with their requirements, passed down to 24 the subsystem as requirements. Those designers are charged p_ 25 with the responsibility of defining the optimum way to meet V
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6850 07 02 74 1 DAVbur 1 those.
,c-(-) 2 Finally, we get component design specifications 3 which can go out.
4 Now, throughout all of this, because of the 5 number of organizations that are necessary -- and even 6 within the organizations, all the suborganizations are 7 involved, and the interactions between all of these 8 functions -- it is absolutely necessary on a systemwide 9 basis to be able to have the multi-system interface analysis 10 performed and identified.
11 And so, we go through and we have interface 12 requirements documents which identify how and in which areas 13 the interfaces are to take place, so that all of the
(,,)
14 organizations involved know what is required for them, and 15 other people's requirements are taken into account.
16 For example, a heat transport design engineer 17 cannot establish any requirements for himself. His job is 18 to understand what the cooling requirements are for 19 everybody else in the plant. Then he takes all of those 20 requirements, develops a design envelope to meet all of 21 those, and he makes sure that each of their requirements is 22 met. He is an interface. The heat transport man is an 23 interface. He has the interface with everybody in the 24 plant. ,
1 25 The fuel designer must interface with everybody l
(,~') I s- :
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1 6850 07 03 75 ,
i 1 DAVbur 1 in the plant. I 2 The purpose for the scram rods in a plant is to 3 be able to turn off the heat generation, that is, turn off 4 the reactivity fast enough so the temperatures don't get up 5 high enough to cause damage to the fuel.
6 Well, the performance requirements of the fuel 7 must be translated to the control rod design engineer. It 8 must be translated to the plant protection system designer.
9 So you need to worry about those interfaces, and 10 that is what this will provide.
11 (Slide.)
12 To give an example, most people don't pay 13 attention -- I shouldn't say that, but it is not explicitly O 14 recognized by many people. For example, in safe, economical 15 power if you want to maintain operations there are four 16 stages for the owner-operator which represent plant 17 operation, and they are the ones which are listed here.
i 18 For illustrative purposes, we have taken to 19 maintain energy production. This is the one that most 20 people concentrate on. This is when the plant is producing i
21 power.
22 The functions at this stage, the functions that 23 have to be performed, are described right here. All of 24 these functions must be performed during energy production 25 if we are to meet the functional requirements to meet that ACE-FEDERAL REPORTERS, INC.
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6850 07 04 76 1 DAVbur 1 80 percent availability number.
2 Therefore, we allocate among each of these what 3 is their requirements. So, for example, everybody knows 4 about having to produce reactor energy to maintain energy 5 transfer and to convert that energy. But along with that 6 there are many other considerations that have to go in, and
- 7 they are here.
8 We perform a functional description for all of f
i 9 those and define what the functional requirements are. Now, 10 none of this is any different than people do it right now.
11 Now, I would like to give you an example of how
- 12 the design evolves.
13 (Slide.)
O 14 In this viewgraph, I am talking about how to get 15 a goal one design, as we call it; that is, maintain normal 16 operations.
17 The first thing we do is we develop the functions 18 and requirements and make design selections to meet the goal 19 one requirements to maintain plant normal operations during 20 each one of those stages.
21 (Slide.)
22 Pictorially it looks like this. Here is goal 23 one: maintain normal operations. We have requirements for 24 that.
25 Those requirements then are allocated down to the ACE FEDERAL REPORTERS, INC.
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6850 07 05 77 1 DAVbur 1 functions. Taking these requirements, then, we perform Based on this, we provide 2 these analyses and trade studies.
3 a design selection, and from that you get the design.
4 DR. CARBON: What is a trade study?
5 MR. MILLUNZI: For example, you can perform what 6 kind of a vessel should we use in the circulator, for 7 example. Should it be a water-cooled circulator, a magnetic 8 circulator in the cooling?
9 All kinds of different studies.
i 10 DR. CARBON: You mean tradeoffs?
11 MR. MILLUNZI: Tradeoffs.
12 DR. CARBON: All right.
13 (Slide.)
O 14 MR. MILLUNZI: Then how do we proceed on to 15 providing the design that not only can meet the goal one 16 requirements but also now can meet the goal two '
17 requirements?
18 Okay, the way we do that, then, is you take that 19 goal one design that you developed, which was focused on q
20 developing to meet the goal one requirements. Then we 21 develop the goal two functions and requirements to maintain 22 plant protection.
23 This is just independent. We just said, okay, 24 our goal is to maintain plant protection. What are the 25 requirements? What are the functions that we need to ACE-FEDERAL REPORTERS, INC.
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6850 07 06 78 1 DAVbur 1 perform to do that.
2 Then we evaluate the goal one design, the 3 existing design, to see if they meet the goal two functions 4 and requirements. If not, we develop the modifications or t
5 design changes.
6 (Slide.)
7 This is illustrated in this one here.
8 We start off. We have developed this plant 9 design. Then we take a look at the goal two requirements 10 and the functions and the requirements independently. Then 11 we evaluate to see if these selections meet the goal two 12 requirements, if that plant as it presently stands meets the 13 goal two requirements. i O 14 If the answer is "yes," we then proceed to 15 develop a design which will meet goal one, goal two, and 16 goal three requirements.
l 17 If the answer is "no," then we have to determine
( 18 what the additional functions and/or additional requirements 1
19 are.
1 20 You go through the same analytical loop, but now 21 you have to keep in mind that if I make a change to meet the 22 goal two requirements, am I affecting its ability to meet 23 the goal one requirements? You have got to have this 24 integrated. You have to keep looking, balancing and trading 25 off and keeping all requirements in mind.
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i l
6850 07 07 79 j 1 DAVbur 1 So do we go through this analysis with these I l
\ 2 design selections? That means that the plant will now meet i 3 both the goal one requirements and the goal two 4 requirements, and now we take that design, repeat that 5 process to see if it meets the goal three requirements.
6 I now come to what I always describe as my 7 favorite viewgraph.
8 (Slide.)
9 It is this one here. On one viewgraph I think we 10 have described the nuclear industry.
11 What we have up here is what has to be done.
12 What we have here is how we do it. When we put 13 the functional requirements on here, we describe how well it O 14 has to be done.
15 Having identified the' product and how we want to 16 get there, we then turn to the four institutions to provide 17 that product.
18 We have a design organization, a construction 19 organization, an operations organization -- this is "o" --
20 and a maintenance. Big "O" is operations and maintenance.
21 It is important -- okay -- therefore that from 22 the beginning all institutions must understand what is 23 required from them from the outset.
24 Up here, the end-user has to understand and l
l l 25 determine what it is he wants.
r ACE-FEDERAL REPORTERS, INC.
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l 6850 07 08 80 1 DAVbur 1 These people have to understand that is what he 3 Now, the design organization is in here. He must 4 understand when he is going to design a product which is goi 5 to meet this requirement -- he has to have an understanding 6 of how the constructor intends to build his design, and he 7 must understand what the operations and how the operations 8 intend to use his product.
9 Also, the constructor must know what the designer 10 had in mind so that he can maintain that intent in his 11 construction practices, and he also must know that that 12 device that he built -- how will it affect the operations 13 and maintenance.
O 14 You do not want to have a Chevette situation, 15 where I build an engine and if I want to change the spark 16 plug I have to pull out the whole engine.
17 So from the very beginning he has to know.
18 Also, the people who are going to operate-and 19 maintain it must understand how this guy's design, or the 20 design he intended to meet these, and how he expected them 21 to operate and maintain it. He also has to provide feedback 22 so that the construction features and practices meet his.
23 Therefore, you have to bring all of these 24 together in the beginning, and you have to integrate all of 25 them, and this demands an integrated approach, and that is ACE-FEDERAL REPORTERS, INC.
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Y
! 6850 07 09 81 1 DAVbur 1 exactly what we have.
2 We have the end-users, we have the NSSS, we have 3 the constructors, and we are interacting with the 4 operations. From the very start ws are using this.
5 We provide a format which clearly defines what 6 everybody is supposed to do. We have a format which defines 7 what all the interfaces are, and we have that documentation 8 logic which will maintain the visibility as people come and
- 9 leave the program.
10 And that is a brief discussion of how we do it 11 and what we do, and I will emphasize over and over again 12 there is really nothing very sophisticated.
' 13 DR. CARBON: Fine.
O 14 (Slide.)
15 MR. MILLUNZI: I think now you have sat through 16 all this approach and what it is, and I think it is 17 appropriate at this time, I guess, for you to see what all 18 of this has wrought.
19 And so, Bill Sheridan, from Stone & Webster, will a
20 start off talking about the overall plant. Tony Neylan will 21 then follow with the NSSS.
22 DR. CARBON: Mr. Sheridan, we will need a break i
23 here pretty soon. Is it appropriate to interrupt yours?
i 24 MR. SHERIDAN: I think so. You mean to break?
i 4
25 DR. CARBON: If it is all right to interrupt it, ACE-FEDERAL REPORTERS, INC.
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,,-w-w , v , . - ------- a-n.- an~,, , - - - -
6850 07 10 82 1 DAVbur 1 go ahead for a while. If you had rather, we will take a
~n) k- 2 break now and not interrupt you.
3 MR. SHERIDAN: I will take approximately 15 to 20 4 minutes, and Tony Neylan, who should follow in sequence with 5 me right away, another 15 or 20 minutes. So it will
)
6 probably be near 11:30 before the next time to break.
7 So either right now would be the natural point to 8 break or approximately 11:30.
9 DR. CARBON: Let's take a break now.
10 (Recess.)
11 DR. CARBON: Let's go ahead, Mr. Sheridan.
12 (Slide.)
13 MR. SHERIDAN: My name is Bill Sheridan, Project O
V 14 Manager from Stone & Webster Engineering Corporation.
15 I will address plant design in the. overview sense 16 for the modular HTGR.
17 (Slide.)
i 18 The outline I will follow to indicate some of the 19 key design selections that we have made as a result of going 20 through the functional analysis, the integrated approach 21 that Andy has previously described, then to take a look at 22 our current plot plan, including the nuclear island and the 23 energy conversion area, and then look in more focus in the 24 nuclear island on the cutaway view of the reactor in the 25 reactor building, in its position on the site.
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i-6850 07 11 83
- 1 DAVbur 1 Then looking at the energy conversion area in i
p v 2 terms of a typical -- or the flow diagram, which is i
3 representative of a typical HTGR plant, rather than looking i
4 at the hardware. But look at the flow diagram and the 5 different connections between systems in that sense.
6 That is briefly the outline.
7 DR. MARK: Will you tell us what you think the 8 relative costs of the nuclear island might be compared to 9 the energy conversion facility?
10 MR. SHERIDAN: Can I come back to that?
11 DR. MARK: Yes. I am just asking will you tell
, 12 us that later?
l 13 MR. SHERIDAN: I will present some information 14 related to that.
15 (Slide.)
16 Overall plant key design selections, pursuant to 17 and in compliance with the integrated approach that we have 18 discussed previously.
19 First of all, we have a four-module plant with 20 350 megawatts thermal each module. So there's four modules 21 arranged tandemly producing approximately 600 megawatts 22 output, 558 megawatts electric.
23 DR. CARBON: This is presumably based on 24 economics?
25 MR. SHERIDAN: This is based on economics. It is ACE-FEDERAL REPORTERS, INC.
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1 1
3 6850 07 12 84
, 1 DAVbur 1 based upon operations, goal one. It is based upon goal 2 two. Goal zero of course is economics. And then it does 3 meet the safety goals.
4 4 DR. CARBON: And the fact that you have four a
5 modular units rather than three or five or something is 4
6 economics?
7 MR. SHERIDAN: Yes. Why four?
8 one, it represents that which the utility user 9 perceives as his type of preferred configuration, that load 10 level or that level of output. The 4- to 600-megawatt 11 electric range is right now the preferred range.
4 12 One of the flexibilities with this modular setup 13 and in the construction of it, you could operate one or two i
O 14 or three or four. So you have that flexibility.
i
~
15 This encompasses, this envelopes that
- 16 consideration.
i 17 1
18 19 20 21 22 j 23 24 25 i
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1 DAV/bc 1 The next significant selection made was two 300
! 2 megawatt electric turbine generators that provide the load l 3 for the four modeles.
i 4 The next key selection is a single control room, 5 which controls virtually automatically all of the modules in 6 the two turbine generators.
7 DR. CA5 BON: We did understand correctly you 8 could run one turbine on one module at 150 megawatts 9 electric?
10 MR. SHERIDAN: That's correct. We can do that.
11 And I think we'll see that when we get into the flow diagram 12 of the energy conversion area.
13 The last key selection is the separation of the O 14 nuclear island, which contains the modules and associated 15 equipment and the energy conversion area which contains the 16 turbine and its associated equipment.
17 So there's a distinct separation between the h
i 18 nuclear island and the conversion area.
19 Looking at the next viewgraph, you have a foldout 20 in the handout which shows the complete plot plan. What I'd 21 like to do is to build that plot plan up in stages.
22 (Slide.)
23 Starting with the outline of the yard itself, the 24 site itself, it's divided into two separate segments, the 25 energy conversion area and the nuclear island. The nuclear ACE-FEDERAL REPORTERS, INC.
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8650 08 02 86 f 1 DAV/bc 1 island is bound by the double security fence completely 2 around it. .
I 3 The energy conversion area has a normal type of 4 fence boundary for that type of facility. The dimensions, 5 we're talking approximately 900 feet in the north-south 6 direction; the nuclear island east-west direction is 7 approximately 600 feet; and another 900 feet in the 8 east-west direction for the energy conversion area.
9 Those are approximate dimensions that may be of 10 interest. The total land area less the switchyard is i 11 something on the order of 26 acres. The total land area out 12 to the exclusion area boundary of 425 meters is 13 approximately 140 acres.
( 14 So 26 acres for the site proper, 140 acres out to 15 the site boundary, dimensions roughly 1,500 feet by 900 16 feet, including the switchyard.
17 Now we build up the plot. First, is the heart of l 18 each area outlined in red. The heart of the nuclear island 19 of course is the core modules arranged in a north-south i
20 configuration. At the head of the tandom configuration is i
21 the reactor service building.
22 Now, the number isn't on here, but it's on your 23 handout. That is item 4 on the plot plan that you have.
24 Over in the energy conversion area is the curbine building.
, 25 DR. SIESS: Excuse me. Before you leave that i
l i
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8650 08 03 87 1 DAV/bc 1 area, what I had on my handout shows a standby power
! - 3
/ 2 building, which I assume is diesels.
3 MR. SHERIDAN: Yes.
4 DR. SIESS: There's a notable lack of separation 5 there.
6 MR. SHERIDAN: Item 9 is the standby power i
7 building. j l
4 8 DR. SIESS: I assume there's two diesels?
9 MR. SHERIDAN: At the present time, it is ,
10 contemplated to have two generators.
11 DR. SIESS: And they're side by side?
4 12 MR. SHERIDAN: Approximately side by side.
j 13 DR. SIESS: Have you considered sabotage in this i 14 plant design and layout?
15 MR. SHERIDAN: Sabotage is a feature that derives i
16 from I think virtually every one of the goals that we have.
17 There are some specific features of the entire plant that
, 18 enhance its sabotage resistance.
19 First, if you look at the dimensions of the 20 sensitive area, the nuclear island area, it has been
! 21 shrunken down. It's not the entire site. It's just one 22 section of the site.
23 So your sensitive area has been minimized. Item l 24 2, each one of these reactor modules is embedded in the 25 ground such that the top of the reactor building is at ACE-FEDERAL REPORTERS, INC.
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W 8650 08 04 88 1 DAV/bc 1 grade level. And we'll see that in the cutaway as we go 2 o ". . That is an enhancement against security risks and 3 sabotage.
4 Another way of looking at this is, within the 5 island itself, you have dispursal. These four reactors are 6 dispursed within the island. So that the concept of 7 disbursal is an enhancement against sabotage.
8 DR. SIESS: That was just my point. It doesn't 9 seem to apply to your emergency power.
10 MR. KELLEY: I think you should clarify that we 11 have not identified a need for safety-related emergency 12 power systems, such as diesels or gas turbines, at this 13 time.
14 DR. SIESS: You don't need them?
15 MR. KELLEY: Right now, we don't believe that 16 these would be safety-related. These would be here for 17 availability and invec.ument protection reasons.
18 DR. FiESS: Public health and safety doesn't 19 require them?
20 MR. KELLEY: Public health and safety, we do not 21 believe require diesels.
l 22 DR. SIESS: Nor does your 10 to the minus 5 on 23 serious damage?
24 MR. KELLEY: From an investment protection
( ,
_ 25 viewpoint, we may need that to protect some of the equipment 1
- )
s_-
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t 8650 08 05 89
. 1 DAV/bc 1 in the plant. But, from a public health and safety view in 2 terms of meeting those top level regulatory criteria, we 4
! 3 believe they're not necessary.
4 DR. SIESS: Okay.
s 5 MR. SHERIDAN: That was my next point, Archie.
6 Moving on, rounding out some of the more 7 important buildings on the nuclear island, we have number 5, 8 the control building. Adjacent is the personnel services 9 building and the waste management building, the rad waste 10 building. And here we show item 9, the standby generators, 11 which, as Archie said, are not safety-related based upon 12 the functions and requirements.
13 That's not to say we don't have safety-related
( 14 loads, but they would be handled separately through an f 15 uninterruptable power supply by batteries.
16 We do recognize there are class one type loads.
i 17 DR. SIESS: I assume the batteries are separated.
18 MR. SHERIDAN: The batteries are separated and 19 the batteries are in the reactor service buildings, which 20 would be -- we'll see them on the next escalation of the 21 buildup, here, here, here and here.
22 This completes the final buildup of the plot T
23 plan. We have inserted items 2 and 3, auxilliary buildings,
,i 24 reactor auxilliary buildings. One set for the first two 25 reactors and then a set down here, east and west reactor ACE-FEDERAL REPORTERS, INC.
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8650 08 06 90 4
1 DAV/bc 1 auxilliary buildings for the third and fourth reactors.
2 Item 14, helium storage, pumping. Item 10, fuel 3 for the standby generators. And item 8 is air-blast heat 4 exchangers. We do have separate warehousing for the energy 5 conversion area and the nuclear island.
l 6 Thirty-two is for the nuclear island; 33 for the 7 energy conversion area. The standard buildings comprise the 8 remainder of the energy conversion area; 25 the switchyard; l
9 30 the fire pump house and water storage tanks; the 10 configuration of 15, 16 and 17 is the auxilliary boiler 11 maintenance building and water treatment building.
12 Item 21, nonessential switch gear. That about 13 rounds out the totality of the plot plan. Now, what we
( 14 would like to do is take a look at one of these reactor 15 buildings. We'll be looking from the south to the north in 16 elevation view. That will be the next viewgraph.
17 Looking from the south to the north...
I 18 DR. MARK: I'm interested that you mentioned 19 south to north. It doesn't really matter, I suppose. I 20 could put it in Arizona and have it go east-west instead.
21 MR. SHERIDAN: Yes. The only reason I did that 22 is when you see this, you'll see the steam generator en the 23 left side of the reactor vessel; and, before, you had seen 4
24 it on the right side of the reactor vessel.
! 25 That's because we're looking from south to north.
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l 8650 08 07 91 1 DAV/bc 1 It's just a matter of orientation.
(/
~ 2 (Slide.) l 1
3 This is really the heart of the nuclear island. l 4 Tony Neylan will get into this in more detail. But I would 5 just like to outline, one, the reactor building which is 6 described -- by the way, this is a buildup in your 7 viewgraph. You have a top enclosure on this; I'll put that 8 up in a minute.
i 9' But the first thing to notice is the reactor 10 building itself, which is outlined in concrete here. That 11 comprises the reactor building, and that houses the steam 12 generator and the reactor vessel.
13 The reactor building is set at grade level. This O 14 is grade level right here. The depth is 151 feet, 15 approximately 151 feet. It's a right circular cylinder on 16 down into the depth, and it has a diameter of approximately 17 63 feet.
18 The reactor building is a concrete building, 19 approximately three feet thick here, two feet thick up-20 there, and it's designed for seismic. Eighty-five percent 21 of the site is designed for tornado and designed for missile 22 protection.
.23 DR. CARBON: For curiosity, what is the seismic?
24 MR. SHERIDAN: Point 3 and .15. That comes right 25 from the utility user's guide. It's specified in there.
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8650 08 03 92 1 DAV/bc 1 DR. MARK: What was the reference to missiles? I
(~)\
\
w 2 don't see any missiles coming in at the 151 foot cap.
3 MR. SHERIDAN: Right here is grade level. You j 4 have something above gro a. And that's one of the features 5 of being embedded, is that the missile protection function 6 is accomplished more easily. It's facilitated. That's one 7 of the advantages.
8 DR. SIESS: Do you or the licensing staff see any 9 problems with determining the seismic design for an embedded 10 structure?
11 MR. SHERIDAN: At the present time?
12 DR. SIESS: Or a seismic analysis, shall I say?
i 13 MR.' SHERIDAN: I'm going to ask Tony Sweeney from
() 14 Bechtel to address that. They are doing the seismic 15 analysis.
16 DR. SIESS: I'm sure they're doing it but I don't 17 believe the staff has licensed an embedded structure in 18 recent history. Have they explored with the staff what kind 19 of questions on soil structure interaction?
20 MR. SWEENEY: If I understand the question 21 correctly, yes. We are exploring all the likely questions 22 that go into the soil structure interaction, and we're doing 23 an analysis to look at that.
24 We understand it is an area where I guess we have 25 to develop an understanding with NRC about how this will be ACE-FEDERAL REPORTERS, INC.
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8650 08 09 93 1 DAV/bc 1 handled. So we're sort of using the state of the art, you 2 know. The evolving ways of addressing this and the computer 3 codes to develop that area.
4 DR. SIESS: And you are working with the staff 5 people on that?
6 MR. SWEENEY: In the future, we will. I can't l 7 that, at this point, we have given them our analyses but our 8 analyses are in progress. And, at the appropriate time, 9 when we have completed our position, then we will interact 10 with them.
11 DR. MARK: Suppose it's rather soggy where the 12 site is chosen for water leaking into the minus 150 foot 13 level?
()
14 MR. SWEENEY: If I might address that, first of
- 15 all, we have a criterion, a user requirement, to be able to
- - 16 site the plant on 85 percent of sites that are available.
17 So, first of all, there are some sites that 18 aren' t acceptable for any nuclear and may not as well be 19 acceptable for this plant, j 20 We do believe that our water ingress control will 21 sufficiently control water ingress and that shouldn't be a 22 problem for the structure.
23 DR. SIESS: I don't see a sump.
7
! 24 MR. SWEENEY: There will be. All the things you i 25 would expect to be in an embedded structure will be there --
l ACE-FEDERAL REPORTERS, INC.
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8650 00 10 94 1 DAV/bc 1 for water control.
10
() 2[ MR. SHERIDAN: Some other dimensions of 3 interest. The steam generator, some 85 feet tall by 15 feet 4 on top, is the circulator.
5 DR. SIESS: Is that helical tube?
6 MR. SHERIDAN: Helical tube, yes. The reactor 7 vessel is approximately 75 feet by 22 feet. Tony Neylan 8 will go into that in more detail.
9 Now, I forgot. To complete the building...
10 (Slide.)
11 ...the top of the reactor building, we have the 12 maintenance bay building. This is a steel structure that is 13 on top of the reactors and it goes completely from north to 14 south, south to north. It encompasses all of that stretch 15 of area covered by the four reactors.
16 DR. SIESS: How much of what's on the slide is 17 safety-related?
18 MR. SHERIDAN: How much on this total viewgraph l
19 is safety-related? I'm going to let Fred Silady address f
20 that question. Fred...
21 DR. SIESS: It might be easier to say how much is 22 not safety-related.
23 MR. SILADY: Probably, the best way to approach 24 the question would be to go through the licensing approach l 25 and talk about the process by which we term things
() l ACE-FEDERAL REPORTERS, INC.
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]L DAV/bc 1 safety-related and then, at the end of that, tell you the 1
3 DR. SIESS: Assume that I know that, because I 4 do.
5 MR. SILADY: Roughly, the fuel, the vessel, the 6 graphite core.
7 DR. SIESS: Everything on the righthand side 8 here?
i 9 MR. SILADY: I wouldn' t go so far as to say
. 10 everything on the righthand side. And even on those things 11 that I'm mentioning, it is a particular function that will 12 be safety-related.
13 For example, on the vessel, it will be a function
( 14 to control chemical attack. It may not be a function to 15 retain primary coolant. So, as we go through the 16 presentation that's coming up here in a few minutes, we'll 17 get into a little more depth on that.
18 DR. SIESS: I'll wait.
19 MR. SHERIDAN: But in the maintenance bay 20 building is a 150-ton bridge crane. It runs up and down 21 from the reactor sites. In_the power conversion area, the 22 turbine building is a standard turbine building, 23 approximately 200 feet by 200 feet.
24 I think it's of interest to look at the turbine 25 cycle in terms of a simplified flow diagram.
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8650 08 12 96 1 DAV/bc 1 (Slide.)
2 There are four steam generators producing the 3 steam at generally fossil-type of steam conditions. That 4 is, 1,000 degrees Fahrenheit, 2,500 pounds pressure. That's 5 the output of the four steam generators. They are headed 6 together but isolable.
7 Normal operation would have them headered. Off 8 that header come the two turbine generators with the HP/IP.
9 That's a little on the spot correction. That should be IP 10 instead of LP'in that middle bank of turbines there, 11 Intermediate pressure, and then a low pressure turbine to 12 the generator, exhausting at 2.5 inches of mere'ury to the 13 condenser. The condensate pump feedwater heaters. The de-O 14 aerating feed tank feedpumps on into a feed header, which is
- 15 combined.
16 In other words, the feed is headered and the i
17 steam is headered. It will have the capability of isolation 18 of any of the steam generators and ability to provide steam 19 from a steam generator to a turbine.
l 20 So there will be a valving arrangement to provide 21 that type of flexibility. That's the significant part I
! 22 think of this viewgraph, is the headering of the steam and 23 the feed. The rest is rather standard power plant.
l 24 (Slide.)
25 In closing the overall plant parameters, the t
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_~ - .. - . . ..
8650 08 13 97 1 DAV/bc 1 entire plant, some of the larger parameters, significant 2 parameters, are core power, which is of course four times 3 350 to 1,400 megawatts thermal.
i 4 The gross power to the turbine is slightly above i
5 that due to the circulator heat input and minus the ambient
- 6 losses, and some other losses that we can get into. Steam 7 pressure, conventional; fossil parameters so to speak, 1,000 8 degrees Fahrenheit; 2,415 psia exhaust, rather standard f
9 exhaust pressure; two and a half inches of mercury, total i 10 generation 600; auxilliary power requires some 42 megawatts, 11 with a net generation of 558 and an overall plant efficiency 12 of approximately 40 percent.
! 13 And that's the total plant layout and some of the O 14 important parameters.
.. 15 DR. MARK: Now, the numbers on that slide,.I i
16 believe you have said were chosen as the result of a survey i
! 17 of the utility representatives. That is, this is about a l
18 good size. This modular feature is good.
19 Is that mainly influenced by U.S. consumers, i
20 utilities? Or is it also with a view ~to outside the country I
21 interests?
22 23
! 24 l
C:)
l l
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l 6850 09 01 98 1 DAVbw 1 MR. SHERIDAN: The requirements for the
~
2 particular thing?
3 DR. MARK: 600 megawatts.
4 MR. SHERIDAN: That is from U.S. user utility 5 requirements. It so happens that this plant does have the 6 feature of relying on a single monitor or two monitors which 7 could be sold elsewhere, but the totality of these 8 parameters really derives from U.S. user requirements.
9 DR. MARK: Now also the question you showed us in
- 10 one of the early slides, you're securing area, the extra 11 barbed wire and stuff, enclosed something, but didn't i
12 enclose everything. I was wondering why are the diesel 13 geneators outside inside of inside that barbed wire?
O 14 MR. SHERIDAN: Let me put the Vugraph back on.
15 The barbed wire doesn't really show up very j 16 well.
17 DR. MARK: No, I hope not.
18 DR. CARBON: They're inside, Carson.
19 MR. SHERIDAN: What I said is, the barbed wire 20 markings don't show up very well. It's difficult to see 21 them.
22 DR. MARK: Where are the diesels then?
i 23 MR. SHERIDAN: The standby generators are located 24 here, you see in outline form.
s 25 DR. MARK: So my wild-eyed saboteur can't get at ACE-FEDERAL REPORTERS, INC.
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6850 09 02 99 1 DAVbw 1 them.
' 2 MR. SHERIDAN: Let's make it clear --
3 DR. CARBON: I think it's clear now, and maybe 4 we'd better move ahead.
i
. 5 DR. MARK: Now you were going to tell me about 1
6 the relative cost. -
7 MR. SHERIDAN: I was going to tell you about the 8 relative cost. The total cost of the plant is about $1 9 billion, not including AFDUC. I think Archie Kelley has 10 some figures on the division of them. We are working on the 11 division between the nuclear island and energy conversion 12 areas. I think Archie has some of those figures, but that
, 13 is a feature in our cost estimate. We want to be able to 14 segregate those things and be able to push a button and say, 15 here's what this is on the nuclear island, so we're working 16 towards that.
]
- 17 Archie, do we have any initial figures on that 18 yet?
l 19 MR. KELLEY: The split is roughly 700 nuclear i
1 20 island, 300, including development costs.
! 21 MR. SHERIDAN: 700 nuclear island, 300 energy 22 conversion, including development costs.
i i 23 DR. MARK: I was asking about it, because that I
l 24 300 is pretty independent as to whether it w'"s nt-lear or -
25 nonnuclear.
! ()
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6850 09 03 100 l
2 DR. R an u.
3 MR. SHERIDAN: Going back to the road map, Tony 4 Neylan will now discuss the nuclear island and really focus 5 in on the NSSS, and we are here on the road map discussing 6 plant design.
7 (Slide.)
8 MR. NEYLAN: Good morning, gentlemen. I'm Tony 9 Neylan with GA. I am a division director responsible for 10 the HTGR programs under the DOE contracts, and I'm going to 11 give you a brief overviw. The handouts that you have are in 12 a slightly different order. I will be trying to catch up a 13 little bit of time by not completely reading through of the O. 14 text ones that are on there.
15 (Slide.)
16 I will also be coming back later after the 17 licensing approach to talk about the nuclear steam supply in 18 more detail. Specifically, I will be at that time 19 addressing each one of the nuclear steam supply systems that 20 are identified on this Vugraph, which identifies those 21 systems which comprise the nuclear steam supply scope with 22 the exception, in fact, of the reactor cavity cooling 23 system, which was on the last presentation, in fact, as part 24 of the nuclear island in the balance of plant.
25 (Slide.)
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6850 09 04 101 1 DAVbw 1 In this overview, I will concentrate on the
-m 1
(/
's- 2 configuration and design parameters, and first of all, which 3 in slightly out of order in your handout --
4 (Slide.)
5 -- addressing the design parameters. And this is 6 one module. I think that's about four sheets in there. One 7 module of the overall heat balance diagram that Bill 8 Sheridan showed you, and it indicates on here a single 9 reactor core generating 350 megawatts thermal, when supplied 10 with helium at approximately 500 degrees and delivering an 11 average temperature of 1268 degrees F from the core. The 12 nominal pressure of the primary circuit is 1925 psia.
13 The steam generator will deliver 352 megawatts
(~s)
'~'
14 thermal when supplied with cooling water at 380 degrees F, 15 3000 psi, and supply steam at 1005, 2500 psia. These 16 numbers should be consistent with the numbers shown on the 17 larger diagram. Of course, this then goes into the header 18 system and the multiplier reactor modules to the twin 19 turbines. The helium is pumped through the primary circuit 20 with a helium circulator, which develops a total head of 21 about 13.2 psid. Those figures are, in fact, the key ones.
22 (slide.)
23 They are shown on this next sheet in there. I 24 won't read them thorugh for you again. You may note that, 25 in fact, the outlet temperature of this HTGR is less than
(~)
s-1 l
l l
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6850 09 05 102 1 DAVbw 1 Fort St. Vrain, which delivers about 1400 degrees F average 2 core outlet temperature.
i 3 DR. MARK: Could you make just a rough comment on 4 a couple of vague questions?
5 Supposing we found ourselves using plants of this 6 sort. The helium requirements and the helium supply, how do 7 those stand out?
8 Also you mentioned a DOE contract. What's the 9 rough size of that contract?
10 MR. NEYLAN: The contract from DOE is part of the 11 HTGR national budget which averages around about $30 million 1
12 per year. The contract to GA is approximately $15 million.
13 The helium supply is a completely contained
' ( 14 system, and therefore, there is an initial fill of helium, 15 and then a small makeup, which is expected to be less than 1 16 percent of the inventory per year, including all helium 17 leakages, and so forth.
18 DR. MARK: But our national supply of helium is i
19 finite and has even been reduced since 10 years ago.
i 20 MR. NEYLAN: Based on the projected market 10 21 years ago, there was a lot larger plant with a lot greater 22 requirement for helium. There is perceived to be no helium 23 inventory supply problem, and there still appear not to be, 24 even with the large economy of these kind of reactors.
25 DR. MARK: Thank you.
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6850 09 06 103 1 DAVbw 1 (Slide.)
2 MR. NEYLAN: I'd now like to talk about the i 3 configuration. You saw this diagram in a reactor structure 4 that was embedded below grade. The features of this are i
i 5 that the reactor core is in a separate reactor vessel from i
6 the steam generator. The reactor vessel and the steam 7 generator vessel represent existing lightwater technology.
8 They use the same materials and are approximately the same 9 size and somewhat lesser weight than actual vessels which 10 have been supplied for the PWR industry.
11 The diameter of the reactor vessel is 24 feet --
12 22 feet, pardon me. The height is about 72 feet and the i 13 control rods stand out a further six feet. The steam 14 generator vessel is approximately 14 feet in diameter and 83 15 feet high. They are connected by a coaxial cross duct of j 16 approximately six feet in diameter.
17 DR. SIESS: That vessel's about the size of a BWR 18 vessel??
i 19 MR. NEYLAN: I will this alternoon show you a l 20 diagram which puts them on one sheet and compares them for
{ 21 you.
22 DR. MARK: Questions that we are often plagued l 23 stress corrosion cracking, do not apply to this vessel at l 24 all, I suppose.
l 25 MR. NEYLAN: In terms of the' water environment t
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6850 09 07 104 1 DAVbw 1 versus helium, that is true 3 generator, then you've got the old stuff back.
4 MR. NEYLAN: I will again this afternoon show you 5 some details of that. It is a helically coiled steam 6 generator with water inside the tubes, and the outside of 7 the tubes is a helium environment. As you'll see, in fact, 8 on the next schematic.
9 I'll go through the flow diagra.m and show you 10 where the helium is.
11 The steam generator vessel is located below the 12 level of the reactor core, which in subsequent potential 13 pressured heat-up events or cool-down events, will preclude O 14 potential natural convection into this vessel and thereby 15 protect the steam generator. The main circulator is located 16 in the cold leg at the top of the steam generator. It is a 17 submerged electric drive motor.
18 Located below here is an independent shutdown 19 heat exchanger with its own circulator. I will come up to 20 each one of those and describe them in a little more 21 detail. Those statements --
22 DR. SIESS: Tony, you'll talk about the bearing 23 lubrication system?
24 MR. NEYLAN: This afternoon.
25 DR. CARBON: Is the Fort St. Vrain steam ACE-FEDERAL REPORTERS, INC.
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6850 09 08 105
, 2 DAVbw 1 generator down or up?
1
. 2 MR. NEYLAN: The Fort St. Vrain steam generator
! 3 is in a single cavity vessel located beneath the core.
4 (Slide.)
] 5 The statements that I just made about the t
j 6 locations and so forth are.on another sheet in your 7 handout. I won't spend time on that.
- 8 (Slide.)
l 9 The next Vugraph traces the helium flow during 10 normal operation, and starting with the cold helium from the J
- 11 circulator, it flows on the outside of the coaxial duct. It 12 flows into the outer areas of the reactor vessel and, in 13 fact, is channeled through to the top of the core. It flows O 14 down through the core, as in Fort St. Vrain, across the hot 15 duct, into the steam generator, where it flows down from 16 the steam generator.
17 The steam generator is an uphill boiling, once 18 through, steam generator, around and back through to the 19 circulator. Feedwater is in at the bottom, helically wound 20 once through and then out on the side.
21 If we can focus in for a moment on the core.
22 (Slide.)
23 The fuel components rely on the basic fuel 24 particle technology that has been developed and demonstrated 25 on Fort St. Vrain and is evolving on the low enriched ACE FEDERAL REPORTERS, INC.
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'6850 09 09 106 1_DAVbw 1 fuels, at this point, to further develop this technology.
2 The basic fissile and fertile particles. The.
3 fissile particle is U 235. You can ignore this, because you 4 can' t have an opportunity to recycle it, if you ever wanted
- 5 to.
6 In this reactor, the current fuel cycle is once 7 through, then out into permanent storage. It's mixed with l .
i 8 the fertile particle into a fuel rod like at Fort St. Vrain, 9 that's approximately a quarter inch in diameter, two inches
! 10 high. These fuel elements are located in fuel holes in a 1
! 11 graphite block next to coolant holes. The coolant flow is 12 down through the block. These block are approximately 14 s
l 13 in hes across, flat, 32 inches high. They are the same as 4
O 14 Port St. Vrain. They're stacked one on top of another in a j 15 hexagonal arrangements.
16 DR. MARK: You mentioned U 235 and U 233. Is 17 plutonium not an alternate possibility?
18 MR. NEYLAN: As a plutonium burner, it is, but'it 19 is not part of this graphite design.
l 20 DR. MARK: Why is it excluded? Just because of f
21 publicity, or what?
22 MR. NEYLAN: It's basically, a low enriched l
23 uranium cycle har been selected to meet with proliferation i 24 requirements and to meet with the user requirements.
25 Publicity. Proliferation. It's not a
DR. MARK:
i
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l l 6850 09 10 107 1 DAVbw 1 because of mechanical or nuclear.
2 MR. NEYLAN: That's correct. The same reactor 3 has the capability with the same plant design to work on 4 fully enriched cores like Fort St. Vrain, or in fact, on 5 recycled cores with the 233 or even with plutonium.
6 DR. MARK: You say fertile thorium. You could 7 just as well have fertile uranium.
8 MR. NEYLAN: Yes.
9 DR. SIESS: And you do intend to have the fertile 10 thorium in there?
11 MR. NEYLAN: Yes.
12 DR. SIESS: Just as a contingency of recycling?
13 MR. NEYLAN: Not only for that. It does develop O 14 a two-particle system which could be used for recycle at 15 some future date. It is used for fuel management within the 16 core, in the design of the core.
17 DR. MARK: Now thorium 232 is converted to 18 uranium 233 for recycle. It leaves you with a very messy 19 radiation problem in connection with the fuel propagation, 20 as compared to going from U 238 to plutonium 239. The gamma 21 rays from uranium 233, I guess.
22 MR. NEYLAN: We are not proposing recycle as part 23 of this. This is a low enriched. We do not intend to 24 recycle the generated U 233.
25 DR. SIESS: Do you get energy from it?
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l 1
6850 09 11 108 1 DAVbw 1 MR. NEYLAN: Does anyone have an answer to that?
T'T O 2 MR. SILADY: Yes.
3 DR. SIESS: A significant amount?
4 MR. NEYLAN: I don't think so.
5 DR. MARK: Low enriched? How low? .
6 MR. NEYLAN: 19.9 percent.
7 8
9 10 11 12 13
,q 14 15 16 17 18 19 20 21 22 23 24 25 l O 1 1
1 ACE-FEDERAL REPORTERS, INC.
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6850 10 01 109 1 DAVbur 1 DR. MARK: Would you be better off with 20.3 2 percent?
3 MR. NEYLAN: From a regulatory point of view, 4 no. That gets into safeguards issues.
5 From an economics point of view, 97 percent would 6 be better fully enriched, as on Fort St. Vrain.
7 (Slide.)
8 An important feature of the design is the 9 configuration of the reactor core. This viewgraph shows a 10 planned view of the core.
11 The hexagonal blocks that I showed on the ,
12 previous diagram are arranged in an annular region of 13 essentially three rings of fuel elements, which are patched O 14 in this diagram.
15 The central region are graphite blocks of a l 16 similar size but are not fuel.
1, 17 The outer region also has two rows of fuel i
18 blocks, removable reflector, and there are permanent l 19 reflectors outside of that, all supported within a core 20 barrel from the steel vessel.
21 The selection, as will be apparent later in the t
22 day, is important in governing the temperatures and the 23 fission product releases that result under various cooldown 24 and excursion events.
25 The control rods are located in the outer f
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6850 10 02 110 1 DAVbur 1 reflector. They are 24 in number. There are six located in 4
v 2 the inner graphite columns, and there are independent l
l l
3 reserve shutdown channels, 12 in number, indicated by the 4 black dots on the diagram.
l 5 DR. MARK: This is uranium oxide or uranium 6 metal?
7 MR. NEYLAN: Uranium oxide, oxycarbide.
8 DR. MARK: Supposing the regulatory gurus said, 3
9 okay, you car c a 30 percent uranium if you want, would this j
10 diagram change very much?
11 MR. NEYLAN: The diagram per se would likely not 12 change at all. The loading -- the fuel loading in an 13 individual block might change, though.
O 14 DR. MARK: So you could really use that diagram 15 with a little shifting all the way up to 90 percent i 16 enrichment?
- 17 MR. NEYLAN
- That is correct.
18 DR. CARBON: What is the reserve shutdown system?
4 19 MR. NEYLAN: The reserve shutdown system are 20 pellets of boronated graphite like those on Fort St. Vrain, j 21 which are released from hoppers and go into holes in the 22 fuel blocks.
j 23 DR. CARBON: And what are the boronated pins 24 there?
- 25 MR. NEYLAN: They are provided for shielding to l ()
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6850 10 03 111 1 DAVbur 1 reduce the doses to the steel vessel.
l'~h l kJ 2 DR. CARBON: Are they all the way around? j l
3 MR. NEYLAN: Yes, they are. That is 4 diagrammatically shown.
5 (Slide.)
6 I thought you would be interested in the ways, in 7 this overvies, at least of removing the heat from the core.
8 The normal heat, of course, is removed by the' heat transport 9 system. Decay heat is also removed by the heat transport 10 system in the same flow diagram that I showed you 11 previously.
12 We have two other means of removing decay heat
_ 13 from the core. One is termed the shutdown cooling systems, 14 and on each one of these I will come back here later this 15 afternoon and discuss it in more detail.
16 (Slide.)
17 In the shutdown cooling system which is provided 18 here, essentially for operations and maintenance reasons in 19 the event that the circulator or normal steam generator is 20 not available, which allows you to do maintenance operations 21 more quickly than if you had to wait for the whole system to 22 cool down naturally.
23 Its flow is from the circulator in the outside of 24 the vessel down through the core. Then instead of going
,- 25 across into the steam generator vessel, it goes down through
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Y -
i 6850 10 04 112
-1 DAVbur 1 a small heat exchanger, down into the circulator, and back i
A- 2 through this system.
j 3 So it is here a small maintenance loop provided i
s 4 in the facility.
! 5 (Slide.)
6 The ultimate means of removing decay heat when
- 7 all primary coolant flow is not available to you -- in other
! 8 words, you have lost the heat transfer system, you have lost 9 the shutdown cooling system -- is by conduction and 10 radiation through the vessel to a passive air cooling system i
11 located in the cavity, which was indicated on the general 12 arrangement drawings that Bill Sheridan 'showed.
4 13 And again, I will come back later this afternoon O 14 and show you some details on that.
15 (Slide.)
16 I did, however, want to point out while I am here 17 the resulting temperatures within the primary system and 18 specifically the peak temperatures within the core that 19 result from this radiation conduction mode to the reactor 20 cavity cooling system, the RCCS.
21 And you can see that the peak temperatures are 22 somewhat in excess of 1600 degrees C. Some 30 or so hours, 23 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> into the event, there will be a gradually cooling 24 down.
25 Again, Fred Silady will provide a little more ACE-FEDERAL REPORTERS, INC.
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h 6850 10 05 113 l 1 DAVbur 1 detail on this. But I wanted to put that in perspective 2 with the resulting effect on the fuel particles.
I 3 (Slide.)
1 4 This is data based on the triso-coated UCO 5 particles which were developed a number of years ago, and 6 there is continuing development in this area, but it 7 indicates that the important point is that the fission gas 8 releases with temperature are very low and are expressed 9 here as a failure fraction of the fuel.
10 But clearly, if one is able to maintain 11 temperatures in this kind of range, then the resulting 12 fission gas release which contributes to accident re, leases 13 are very small. That is a key in the inherent safety of O 14 this system.
15 DR. MARK: What is the melting temperature of 16 uranium oxide?
17 MR. NEYLAN: The silicon carbide starts to break 18 down in this kind of temperature range, 2200. That is the 19 breakdown of the fuel particle. That is the primary barrier 20 to release of the fission products.
21 DR. MARK: That is the most temperature sensitive 22 element in that core configuration?
23 MR. NEYLAN: That is correct.
24 DR. MARK: Steel?
25 MR. NEYLAN: There is no steel in the core at 1
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's / ~ 2 We will show you the resulting temperatures on 3 the other core supports, structural components and the 4 vessel, and they are all within acceptable limits.
5 DR. MARK: Graphite is really quite a bit higher, 6 anyway?
7 MR. NEYLAN: That is correct. In fact, it is 8 still increasing in strength at this point.
9 The key, then, to this design is the selection of 10 the fuel, its performance, the configuration of the core, 11 which allows you to limit this temperature and hence fission 12 product release, and the configuration -- the cooling 13 configuration which allows you to get rid of -- dissipate O 14 the decay heat generated in an acceptable and passive way, 15 based on that conceptual design that I have just made.
16 We believe that the inherent characteristics and 17 the passive safety features of this modular HTGR will fully 18 meet those quantified goals and objectives of economics, 19 investment risk aversion, and safety.
20 The key ones to remind you of are the 10 percent 21 economic advantage versus coal, the 10 percent 22 unavailability from an economic and operational point of 23 view, the investment risk, the loss of plant at 10 to the 24 minus 5, and the PAG sheltering guidelines, the 5 rem g_ 25 thyroid, 1 rem whole body, and 5 times 10 to the minus 7.
\-)
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1 6850 10 07 115 1 DAVbur 1 With that, I would like to end this overview and ,
2 come back and give you some of the details this afternoon.
3 DR. MARK: Just before that maybe, Fort St. Vrain i 4 has been plagued with moisture.
a i
j 5 What have you done about that?
i j 6 MR. NEYLAN: It has, indeed. That was relevant j 7 to Dr. Siess' question.
8 We are very concerned about that, and we are 9 looking at the sources and have been actively looking at the 10 sources, the lessons learned on Fort St. Vrain to correct 11 that problem.
12 (Slide.)
13 DR. CARBON: Mr. Silady, anything you can do to 14 gain us some time would be very helpful. We are 50 minutes 15 behind.
16 MR. SILADY: All right.
17 My name is Fred Silady, Manager of Safety and 18 Reliability at GA.
19 The topic of my presentation this morning will be 20 to continue on this master diagram and proceed from the 21 process, the design approach that we have taken, to develop 22 licensing bases that are specific to the flTGR and apply the 23 process to come up with things such as licensing basis 24 events.
25 My presentation is in two parts. There is a very ACE-FEDERAL REPORTERS, INC.
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6850 10 08 116 1 DAVbur 1 brief overview that goes into, in a capsule way, the y- ~
k- 2 development of the licensing bases.
3 (Slide.)
4 Let me just remind you of our licensing approach 5 that was touched on briefly this morning. There are three 6 steps.
7 It is tied to the top level regulatory criteria 8 as a starting point.
9 The second is the area called the bridge, to 10 develop a process to derive licensing bases specific to the 11 modular HTGR.
12 The third is the application of that bridge.
13 We have heard this discussion this morning. I am
)
14 going to give you a brief capsule of the process. Then I am 15 going to go into a longer presentation that goes into more 16 depth in terms of the process and its application step by 17 step.
18 (Slide.)
19 First, a brief overview.
20 We are going to identify the likelihood of events 21 and classify the events into three regions for a comparison 22 against the top level regulatory criteria. This 23 classification of events, using risk assessment, will form 24 the basis then for the licensing basis events.
25 We are going to examine those events tc identify fc3
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1 DAVbur 1 the functions that we are going to rely upon in order to
}
! 2 meet the top level regulatory criteria. These functions r
3 that we rely on are the basis for our principal design 4 criteria.
1 5 The third step is to choose design selections to 6 accomplish those functions that we are relying upon to meet 7 the top level regulatory criteria. This step leads directly 8 into which systems, structures, and components. Design 9 selections are those which should be termed safety related i
10 to corresponding quality assurance.
7 11 DR. MARK: What do you think of when you say top 12 level regulatory criteria? Do you mean Part 100?
i 13 MR. SILADY: It is as we discussed earlier in 14 Kelley's presentation. It is those quantitative doses or 15 risks to the public health or environment that are plant 16 independent, that are top level, that are quantifiable.
17 With that brief overview, let me go into much 18 detail.
19 (Slide.) i 20 The presentation outline is arranged in a fashion 21 where each of the three areas, the three identified 22 licensing bases, is discussed in terms of first defining 23 some terms, then looking at the step-by-step method with an 24 example application as I go through so that you get a better l
~
l 25 handle on exactly what the method is.
l l
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1 1 DAVbur 1 First, I am going to go through the licensing e3
's-) 2 basis events. Then I am going to talk about the functions 3 that we rely upon in order to meet the top level regulatory 4 criteria in 10 CPR 100, then the equipment classification in 5 the same manner, and then a summary.
6 (Slide.)
7 Starting in with the licensing basis events, 8 these are those of f-normal or accident events used for 9 demonstrating the design compliance with the regulatory 10 criteria.
11 Collectively, the licensing basis events are 12 analyzed in risk assessments for demonstrating compliance 13 with the interim safety risk goals.
O 14 We have defined three kinds of licensing basis 15 events, which I will now define one by one.
16 (Slide.)
17 The first of these we call anticipated 18 operational occurrences. These are events that are expected 19 once or more in the plant lifetime. Therefore, these are 20 events whose frequency we would expect to be greater than 21 once in 40 years, or .025 per year.
22 We analyze these events realistically, and they 23 tfpically appear in Chapter 11 of SARs, and they are 24 compared against 10 CPR 50, Appendix I.
,_ 25 So the anticipated operational occurrences are t f t/
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6850 10 11 119 1 DAVbur 1 compared against Appendix I, which was one of our top level n
kJ 2 regulatory criteria.
3 (Slide.)
4 Design basis events we define as events of lower 5 frequency than the anticipated operational occurrences; that 6 is, they are not expected to occur in the lifetime of one 7 plant. However, we do expect them to occur in the lifetime 8 of a large number of plants, say 100 plants.
9 So typically, we think of these events as falling 10 within the range of once in 40 years down to approximately 11 once in 10,000 years, or a 10 to the minus 4th frequency.
12 If we had 100 plants, one would expect a design basis event 13 somewhere in the lifetime of all 100 of those plants.
14 The consequences of design basis events are 15 analyzed conservatively in Chapter 15 of the PSAR, now 16 against 10 CFR 100, another one of our top level regulatory 17 criteria.
18 (Slide.)
19 The third of the three licensing basis events we 20 call emergency planning basis events. These events are 21 still lower frequency than the design basis events. They 22 are not expected to occur even in the lifetime of all 23 standard HTGRs, where now we are taking this to be 100 24 plants. So they have frequency below 10 to the minus 4th.
25 However, we do have a lower frequency for the ACE FEDERAL REPORTERS, INC.
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6850 10 12 120 1 DAVbur 1 emergency planning basis events, and we take this to be 5
,m 2 times 10 to the minus 7th per reactor year for consistency 3 with the interim NRC risk goals, particularly the acute 4 individual risk goal that is specified at 5 times 10 to the 5 minus 7th.
6 The consequences of these are realistically 7 analyzed in risk assessments and in Chapter 13. These 8 emergency planning basis events would be shown to comply 9 with the projected action guideline dose limits at the 10 emergency planning zone.
11 12 13 14 15 16 17 18 19 20 21 22 r
23 24
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50 11 01 121 1 DAV/bc 1 DR. SIESS: Is there any consistency between this 2 and the PAG dose limits of the previous PAG's of part 100?
3 I can't quite picture how you could get to this stage and 4 meet the PAG and not automatically meet part 100 of the 5 DBE's. These events are worse than DBE's and the doses are 6 a fraction of the DBE's.
7 MR. SILADY: The events are lower in frequency.
8 They may or may not have consequences greater than the 9 higher infrequency design basis events. The basis on which 10 emergency plan basis events are evaluated is the emergency 11 planning zone.
() 12 Our user requirement is that that distance be at 13.:i the exclusionary boundary. But, in terms of getting a 14 i license for an HTGR, that distance could vary and could be i
15 at five miles, like it is at Fort St. Vrain, or 10 miles 16 for existing reactors.
17 DR. SIESS: I'm still confused. If I've got a 10 18 to the minus 4 or 10 to the minus 5 event, because 25 rem 19 whole body --
20 MR. SILADY: I have a diagram that may clear this 21 up.
22 DR. SIESS: I'll wait.
23 DR. MARK: In talking about Part 100 PAG's and
(} 24 stuff, all of this is without the use of containment.
25 MR. SILADY: At this point, I'm describing an j
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( ) DAV/bc 1 . approach that is generic, an approach that's going to take 2 the blue part of our master diagram -- that is, the design 3 that we have evolved from the rigorous process that you've 4 heard about, and translate over into licensing bases 5 specific to the modular HTGR.
6 So, yes, specific application of this process may 7 indeed lead to meeting the 10 CFR 100 at the DAB and meeting 8 the regulatory actions at the emergency planning zone 9 without the use of a standard light water reactor 10 containment.
11 The process, however, could be applied to any 1
12 I nuclear power plant. Let me go to the first of many steps, 13 and I hope to be able to come back to Dr. Siess' point to i
14 f clear up any confusion with regard to the relationship of 10 CPR 100 and design basis events to the protective action 15 l 16 quidelines and emergency planning basis events.
, 17 (Slide.)
18 The first two steps in the process are to define 19 three regions on a frequency consequence plot, risk plot.
20 The second step is to compare the risk assessment of the 21 design to that frequency consequence risk plot.
22 Subsequent steps then will take each of those 23 classes of events and choose from those, select from those, 24 ones that would be anticipated operational occurrences
< 25 design basis events, or emergency plannirg basis events.
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( ) DAV/bc 1 The subsequent steps will then also indicate how 2 each of those are evaluated and against which of the top 3 level regulatory criteria.
I 4 But, first, let me show you a typical risk 5 plot...
6 (Slide.)
7 ...with the regions shown according to the 8 definitions we've just gone through.
9 The first region is the anticipated operational 10 occurrence region, where the ordinate is frequency and 11 infrequency per plant year. The abscissa is mean whole body 12 gamma dose at the exclusion area boundary. The slanted line 13 j is appendix I. On the annualized basis of five millirem per 14 year.
I 15 It extends down to once in a lifetime of one 16 plant, which is once in 40 years.
17 The second region we have defined is the design 18 basis region. It starts from that point and extends down to 19 10 to the minus 4th, which corresponds to events that would 20 -be expected to occur in the lifetime of, say, 100 plants but 21 not expected in the lifetime of one plant.
22 The third region starts from that point and goes 23 beyond these design basis events into the emergency planning 24 basis region. It extends down to 5 times 10 to the minus 25 7. In each of these regions there in a top level regulatory ACE-FEDERAL REPORTERS, INC.
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( ) DAV/bc 1 criterion that the events are evaluated against. In this 2 region, it's Appendix I. In this region, it's 10 CFR 100, 3 25 rem that you just referred to, where we have shown a 4 slant in the line in recognition of the fact that, for 5 higher frequency events within this region, one wouldn't 6 design all the way up to 100 percent of Part 100.
7 So this is at 10 percent at this point. And then 8 the third region, the protective action guidelines of one 9 rem that we heard about earlier, isn't shown because this is 10 the dose at the exclusion area boundary. It is our user 11 requirement that all events in here and all events in here 12 be below one rem, so that there need not be any offsite 13 evacuation.
14 g so, if I was plotting user requirements on here 15 in addition to top level regulatory criteria, I would have a 16 line coming straight down. Of course, we're going to have 17 to have agreement if indeed we want to have our exclusion 18 area boundary be our emergency planning zone that all our 19 events are within that region.
20 But it could be that we could license a plant 21 with an emergency planning zone much greater than the EAB.
22 That's the practice of existing plants today, in which case, 23 the dose at the EAB is much greater than that one rem. So 24 the line isn't shown.
J 25 Did that help at all?
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( ) DAV/bc 1 DR. SIESS: That helps but where would the safety 2 goal put you if you knew what it was?
3 MR. SILADY: The safety goal is off the chart.
4 In essence, since I'm only at 100 rem here, and particularly 5 an acute fatality individual risk.
6 DR. SIESS: You take the LD-50 for acute, the i
7 dose that will kill half the people, half those exposed to 8 it? Is it 400 tem, 250 rem?
9 MR. SILADY: Right. There's a function in there 10 in terms of the probability of dying versus dose. And that 11 would be considered.
12 DR. SIESS: And that would be acute fatalities?
I v 13 MR. SILADY: Yes. And, in terms of latent 14 l cancers, similarly.
15 DR. SIESS: It's manrem then. In terms of latent 16 cancers, you have to reintegrate, don't you?
17 MR. SILADY: That's correct. Later, the last 18 step in the licensing basis event methodology, I will show 19 you the results of our preliminary risk assessment summed 20 over all the events, with the conversion of the consequences 21 to latent cancers and compare that to the 2 times 10 to the 22 minus 6 latent cancer individual risks and goals.
23 But I'm getting ahead of myself a little bit.
, 24 I'm still just on step one, making sure that we have a good 25 base here before proceeding, that this is a center or the ACE-FEDERAL REPORTERS, INC.
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() DAV/bc 1 heart of our approach to the licensing basis events, is the 2 definition of these regions.
3 DR. CARBON: How much agreement have you had with 4 the staff on that?
5 MR. SILADY: As discussed earlier today, we have 6 made presentations to them on this approach. We have made 7 the documentation in a form that's being readied within the 8 next month to be sent to them.
9 (Slide.)
10 The second step was to compare a risk assessment 11 to these three regions on the risk plot. These are early 12 results for the 350 plant, and not all the results are 13 plotted, just those that are dominant of the ones that we 14 have examined at this point.
15 And my purpose in showing this is not to go 16 through and define each of these release categories in the 17 event of...but again to show you conceptually the approach.
18 There will be events throughout all three regions. Some 19 will fall into the A00 region, others into the design basis 20 region, and others of lower frequency and perhaps higher 21 consequences, perhaps not, in the emergency planning basis 22 region.
23 Shown on the chart is the mean value in terms of 24 frequency. And the mean value in terms of consequences with 25 uncertainty bands. The uncertainty bands play an important '
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2 DR. SIESS: Another thing that I have to keep in 3 mind when I look at that is that, almost without exception, 4 when the staff reviews somebody else's PRA, the results tend 5 to move upward by a f actor of 1 to 2 orders of magnitude.
6 HR. SILADY: Yes. 'We are aware of that. One 7 thing I should have mentioned at the beginning of the 8 presentation is that this process that I'm describing, of 9 getting from the end nuclear power plant design product over 10 to the licensing stage will be repeated at each stage of 11 the design.
(} 12 i
So our knowledge, our risk assessments at this 13 j point of a conceptual design, may change as we go into
! 14 i preliminary and final design.
15 We are hoping that the characteristics of this 16 concept are such that we wouldn' t anticipato a large 17 difference in opinion between us and the staff.
18 This provides the basis for which we can have 19 discussions in those areas.
20 DR. SIESS: Now if they only move up and don't 21 move to the right, that doesn't really give anybody much of 22 a problem, because these are all so far over.
23 MR. SILADY: That's an excellent point,
() 24 particularly with this concept, which is more a low 25 consequence concept rather than a low frequency concept.
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( ) DAV/bc 1 And if we can show that the safety is transparent 2 with the passive features, we may be able to have less 3 movement right to left in terms of consequences.
4 DR. SIESS: I think that's a key to much of 5 this.
6 MR. SILADY: I agree.
7 (Slide.)
8 The next two steps in the process are to identify 9 those f amilies of events that fall within the A00 region and 10 to identify that rely upon some design selection to keep 11 them within that region, not to violate Appendix I, to 12 identify those as A00's. There are some typos on this O
(_e 13 l chart.
14 Then to evaluate the consequences of those 15 selected AOo's realistically against Appendix I. Let me 16 show you an example of that on the next page.
17 (Slide.)
18 This is very conceptual but it takes one of our 19 real points that we're seeing in our risk assessment so far, 20 which happens to be a primary coolant leak releasino a small i 21 amount of the circulating activity to the environment.
. 22 The point is assessed at being within our goal.
23 However, if it were not for specific design selections that 24 we consciously made to control release, the point could be 25 out here.
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(_ 1 We intentionally have very good fuel quality 2 specified. That makes the circulating activity quite low.
3 So even the uncertainty band doesn' t exceed the Appendix I 4 region goal. But we would take this event and call it an 5 anticipated operational occurrence and evaluate it and 6 report on it in the PSID in Chapter 11.
7 DR. MARK: What is the main component of your 8 circulating activity?
9 MR. SILADY: Primarily, Krypton 85, 88.
10 DR. MARK: It's gas from the fuel elements that 11 might possibly escape, or might possibly get into the 12 helium.
13 MR. SILADY: The circulating activity arises as a 14 result of as manufactured defects in billions of little fuel 15 particles, such that the coatings aren' t properly coated, or 16 a very small fraction.
17 And, in addition, there is during normal
! 18 operation some statistical failure of a very small fraction 19 of those. Those both can release gases into the circulating l 20 activity.
21 (Slide.)
22 The next three steps do an analagous thing in the f
23 design basis region. There's one extra step thrown in to l
24 account for uncertainties. We identify as design basis 25 events those families of events within the design basis i
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()DAV/bc 1 region in much the same manner as we did for the anticipated 2 operational occurrences.
3 However, recognizing that the risk assessments 4 have uncertainties and those uncertainties may cause the ,
5 point to be on the border of the defined regions, we say 6 that if the upper or lower bound frequencies extend into the 7 design basis region, then those too are candidates to be 8 design basis events.
9 And, repeating, these events are evaluated 10 conservatively and shown against 10 CFR 100 and Chapter 15.
11 And there's a page, a couple of pages on gives us an 12 example. I'm going to skip one of the pages in the package, 13 since it comes up again.
14 (Slide.)
15 In light of'the time. And I think we will 16 adequately cover it when we get to it a little bit later.
17 So, here, we have this particular release 18 category family of events and there is a design selection 19 that keeps the point here rather than in an unacceptable 20 region that exceeds 10 CFR 100, 21 So the function of controlling release is 22 accomplished by some design selections. And I'll save that 23 story just for a few more minutes, until wo get into the 21 second and third segments of this presentation.
25 (Slide.)
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( ) DAV/bc 1 The emergency planning basis events. The two
, 2 steps with regard to these, the final of the three kinds of 3 licensing basis events are just to examine that region that 4 is beyond the design basis. Find those events that are 5 consequence dominant, the key ones, to turn those emergency 6 planning basis events and to evaluate those realistically in 7 terms of siting evaluations.
8 And I'm going to skip then the next page in the 9 package, which shows one of those being selected, the one 10 with the higher consequences.
11 (Slide.)
12 Step 10 and the final step then for the licensing 13 basis events is to compare the risk assessment of the design 14 I that was developed with the integrated approach to the 15 interim NRC safety risk goals. That's shown in the next 16 table.
17 (Slide.)
i 18 The first column shows the regions for the events I ,.
19 as they are assessed to lie in -- the anticipated 20 operational occurrence region, the design basis region, the 21 emergency planning basis r egion. These are the dominant 22 events.
23 The list is quite a bit longer, of course. But 24 one takes the frequency in events per plant year, takes 25 those doses -- whether it be whole body, thyroid, or ace FEDERAL. REPORTERS, INC, 202 347 3XO Nationside Cmerge Nn))M6m
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() DAV/bc 1 whatever -- converts it to cancers per event.
2 These two mean values are multiplied together and 3 the risks summed to a value at this point of 2 times 10 to a
4 the minus 8th latent cancers per year, i
5 That's compared to the risk goal of 2 times 10 to 6 the minus 6th. And at this stage, we have a margin at 7 approximately 2 orders of magnitude.
8 Note that the majority of the risk is not coming l
9 from the design basis region, or even from that region 10 beyond or the emergency planning basis region, but from the 11 anticipated operational occurrences.
12 So, here, already, at the conceptual design, we 1
13 j are getting some insights about our design and its risk I
i 14 i envelope.
j 15 The second of three segments of my presentation 16 are the required functions, the functions that we rely upon
! 17 in order to meet 10 CFR 100. And it builds on the selection 18 of the licensing basis events.
19 (Slide.)
20 First, some definitions. These functions are 21 those that are needed to limit the radionuclide retention to 22 meet 10 CPR 100 doses for desiqn basis events. These 23 functions and how we accomplish these functions are the 24 basis for the principal design criteria that we will be back 25 later in the year talking to you about. And of course the ace-FliDERAI, REvonrEns, INC.
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6850 11 13 133 DAV/bc 1 principal design criteria go in Chapter 3 in terms of
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( ) DAVbw 1 MR. MILLUNZI: Fred, if I could on that first 2 one, the word " meet," that means those functions which we l
3 are going to rely upon.
4 MR. SILADY: A very good point. One which I
- 5 think will be brought out in the next slide or two.
6 There are many ways that one could mean 10 CFR 1 7 100, even for a specific design.
8 (Slide.)
9 There are two steps in this selection of 10 functions that we're going to rely upon to moet 10 CPR 100.
11 The first is to identify those functions required to meet 12 the top level regulatory criteria and the user utility 1
% 13 safety requirements.
14 So first, I will try to find all the functions 15 tht have to do with radionuclide retention.
16 (Slide.)
4 17 An example of that i s this segment out of our
?
18 functional analysis tree. That is the Goal 3 portion of the 19 tree, down to four levels. The top box roads " Maintain 20 control of radionuclide roloase." And these two ways are 21 the ways that the plant does that. Control personnel access 22 and control radiation. In turn, there are throo ways to 23 control radiation. One of those is, control radiation from 24 the core. Others are from processos and from storago.
25 There are two things one needs to do, in order to control ACE FliDI!RAL RI!PORTEns, INC.
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( ) DAVbw 1 radiation from the core. Control the direct radiation and 2 control radiation transport. There are four ways, required 3 functions to control radiation transport. You can control 4 transport from the core, in the primary circuit, from the 5 reactor building and from the site, expanding the core on 6 down one more layer.
7 One can retain radionuclides in the fuel
! 8 particles and retain radionuclides in the core graphite down i
9 to this level. These are all the functions that we need in 10 order to accomplish all of the top level regulatory criteria 11 and all of the user requirements relative to not nooding 12 !
evacuation, and so on. But that is just one of the O 13 i
process.
14 DR. SIESS: I asked the question earlier about 15 defense in depth, and I'd like to point out that some 16 people, defense in depth, are the various levels you've got
- 17 here. There are at least three different definitions of 18 defense in depth, but one is in terms of barriers. I think 19 you've got more than four levels here, but I'm not sure.
20 MR. SILADY: One can think of it in terms of 21 barriors, and that's the way the term has evolved in the 22 industry to date, in terms of a quietinq, a vessel, a 23 containment and a site.
24 DR. SIESS: That's one impression of it. The 25 other one is the one that Mr. Millunzi qave.
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( ) DAVbw 1 MR. SILADY: We believe that the one Mr. Millunzi 2 discussed is the way it originally started off, the original 3 definition, and that there are many ways to accomplish a
],
4 defense in depth and, indeed, to meet your top level 5 regulatory criteria, you will see that we are utilizing each 6 of these in meeting the combined top level regulatory 7 criteria and user requirements. Our site is 425 meters. We i
8 have a reactor building. We have a primary coolant 9 boundary, and we certainly have good fuel particles:
10 however, the point I want to make in step 2 is, which of 11 these are we relying upon, in order to meet 10 CFR 100.
P 12 (Slide.)
13 Select those radionuclide retention functions 14 ; required to meet 10 CPR 100 does for design basis events.
15 And your next chart, if you're flipping ahead, 1
16 shows the functions shaded that we need for 10 CPR 100.
17 (Slide.)
18 Based on oiar present plant design and assessment, 1 19 1 we feel that we can show that the sources of radiation and ,
20 radionuclide release from processes and from storage are 01 quite small. Similarly, we feel that our sources within 22 these other areas here are quite small, and I would give you 23 some numbers on those in just a minute. ,
24 So in terms of meeting 10 CFR 100, the key thing O 25 and the only thing we need to do is retain the radionclides i
l Acti FliolinAI. Riii>oRTiins, INC.
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( ) DAVbw 1 in the fuel particles. And from here, I wish to go into a 2 fuel of the design basis events and a few of their thermal 3 and fission product release characteristics and then come 4 back to another sphere of functions that we have found that 5 we need, in order to keep the fission products within the 6 fuel particles.
7 (Slide.)
8 Defore proceeding, I do need to spend just a 9 quick minute, although it's in your package, and I won' t go 10 into it in a lot of depth, to see that first we have to take 11 what 10 CFR 100 is. That is 300 rem thyroid, as an example, 12 or 150 rem at the construction permit and translate that 13 into what are the allowable curies that could be released i
14 8 from the fuel. And we have done this, and we have made 15 assumptions consistent with Reg Guide 1.4 to como up with 16 250 curies of iodine that cut be released from the fuel and 17 without any other consideration of hold up in the vessel or 18 in the reactor building, exactly match 150 rem.
19 For longer term releases, the weather and the 20 I breathing rate and the chi over Os change and the number is 21 higher. Keep those numbers in mind, 250 curies for 22 short-term release and 900 for long-term.
l 23 DR. SIMSS Is there a frequency of probability i
24 thinking that's been assigned to the atmospheric dispersion?
( 25 Is that the average?
l AcsFlintiRAI. RiiPORTliRS, INC.
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( ) DAVbw 1 MR. SILADY: It's directly from Reg Guide 1.4.
2 It's the conservative chi over Os.
3 (Slide.)"
4 This page compaes some of the sources within the 5 NSSS to those corresponding short-term curie releases that 6 go with the 10 CFR 100 of 150 rom. One sees that the 7 circulating activities of the iodines is very small.
8 Similarly, the plateout. The expected activity after 40 j 9 years of operation is a factor of 10 lower, and even our 10 design activity, which is more a 95 percent confidence 11 number is two or three time lower than the 250 curies, which 12 was the short-term release allowable.
4 13 So the purpose of this graph is to say that the 14 , main area of concentration, because of the fuel quality that 1
15 we have that keeps the primary circuit cican, the main area 16 of concentration is in the fuel particles and how to assure I
17 that during an event, those fuel particles remain intact.
18 DR. SIESS: Are the assumptions used in getting
, 19 at those platenut numbers the same as those used at Fort i
! 20 St. Vrain?
21 MR. SILADY: They consider the experience we' eve 22 had at Fort St. Vrain to a large degree. We've seen very 23 good performance of the fuel there.
24 DR. SIESS: llow much of a factor is that, 25 compared with what was assumed in the FSAR for Acti FliniinAI. Riii>onrtins, INC.
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- ( ) DAVbw 1 Fort St. Vrain?
2 MR. SILADY: Actually, I believe in the PSAR for 3 Fort St. Vrain, the plateout per pass was something like 40 4 times, and our experience is now that it is only 1 percent 5 per pass. So those consider the new knowledge and new 6 experience.
4 7 DR. SIESS: A factor of 40 then belov Fort
, 8 St. Vrain?
t 9 MR. SILADY: The actual curies around the circuit 10 on a per megawatt basis are much less. The fuel quality we 11 are specifying for this plant is approximately a factor of 12 5, better than what we have seen at Fort St. Vrain, which is
[
O 13 another probably factor of 10. I've got the backup Vugraph 14 here, better than the specification f rom Port St. Vrain.
15 So we are requiring much tighter fuel quality 16 than at Fort St. Vrain, and we are requiring fuel quality 17 that is even better than what we've seen at Fort St. Vrain.
I 18 DR. SIESS: That gets back to my question. If I 19 look at the column dated design activity platenut --
20 MR. SILADY: The 90 curies.
21 DR. SIESS: If I calculated that on the same 22 basis as the FSAR for Fort St. Vrain, what would I get?
23 MR. SILADY: You would not qet the name number of 24 you did it in terms of per megawatt, okay, because we are 25 specifying fuel quality.
t
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() DAVbw 1 DR. SIESS: I don't know whether you don' t 2 understand the question or don' t want to answer it.
3 MR. SILADY: The former.
4 DR. SIESS: There were certain assumptions made 5 at Fort St. Vrain about fuel quality, not specified, but 6 assumptions, if you wish, or whatever, and plateout, et 7 cetera, that led to some values for plateout. If you made 8 those same assumptions, correcting for megawatts and all the 9 other things, what would you get here instead of 80?
10 MR. SILADY: If you assumed the same assumptions 11 and the same fuel quality and corrected for megawatts, you'd 1 DR. SIESS: You'd still get 80.
14 i MR. SILADY: If you factor *in that we're 15 i specifying better fuel, you get less than 80.
1 16 DR. SIESS: Okay. So 80 would be comparable to 17 l Fort St. Vrain, in terms of assumed fuel quality.
18 MR. SILADY: Excuse me. Maybe I misunderstood.
19 One would not get 80 One would get a number higher than 20 80, if one used the same Fort St. Vrain assumptions, the 21 same fuel quality as Fort St. Vrain and corrected for r
22 megawatts.
23 DR. CARBON: What would you get? About 4007 24 MR. SILADY: I don' t know the number. I know 25 it's at least an order of magnitude. Sorry.
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( ) DAVbw 1 DR. SIESS: So this is taking advantage of state 2 of the art experience.
3 MR. SILADY: That's the point.
4 DR. SIESS: To get down to the 80. Now in the 5 expected activity, do you have a comparable reduction, or is 6 that a realistic comparison with Fort St. Vrain? There have 7 been plateout measurements on Fort St. Vrain; right?
8 MR. SILADY: Both the plateout and the 9 circulating activity has factored in the Fort St. Vrain 10 experience.
11 I think since we've spent so much time on this 12 one point , and since I didn't understand the question at the 13 beginning, let me show yoti very briefly the comparison of 14 the fuel quality of Fort St. Vrain versus the fuel quality 15 that we're specifying hete.
16 (Slide.)
17 This chart show the combined silicon carbide 18 defect and contamination fraction versus essentially time, 19 different cases, but there's an evolution here, and this is 20 the Port St. Vrain specification. Based on the experience 21 that we saw at Fort St. Vrain for the 2240 megawatt plant, 22 the larger plant, we specified the quality factor of 3 l
23 lower.
24 THTR in Germany has this criterion, 50 percent 25 value. AVR has this quality of fuel with the 95 percent
! ace. FEDERAL, REPORTliRS, INC.
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( DAVbw 1 upper bound value.
2 The two HRB designs, 500 megawatt HTR and their 3 100 megawatt, are at this point. Interatom is here. The 4 fuel that Hobeg has made, has made here, here or here.
5 Their criteria in Germany are exactly the same as we are not 6 proposing on a 4 x 350, 6 x 10 to the minus 5 at the 50 7 percent confidence limit.
8 DR. SIESS: That's based on what you've actually 9 been able to produce.
10 MR. SILADY: It's based on experience and what 11 we've been able to produce.
12 I'd like to now then return to the next step in 13 the process --
14 (Slide.)
15 -- which is to examine the design basis events.
16 I am only going to go through one event.
17 DR. CARBON: Could I stop you just a minute.
18 Your last comment was, your fuel specs are based on 19 experience and what you've been able to produce. A lot of 20 that is what you've been able to produce. Itow do you know 21 you've been able to produce it? If I understand what you 22 said, your experience comes from reactor operations, but 23 you've gone well beyond that.
24 MR. SILADY: Yes. There are capsule radiation i 25 tests that are run on the particles and the silicon carbide Acti FliDiiRAl. Rl!PORTI!Rs, INC.
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2 DR. CARBON: Okay. Just a small number. Do you 3 have a statistically significant amount?
4 MR. SI LADY: The base is growing, but there has 5 been considerable interchange between us and the Germans on 6 a joint program exchanging capsules.
7 DR. CARBON: Even so, do you have a statistically 8 significant amount yet? I don' t know how to define that.
9 MR. NEYLAN: The Germans have produced 10 statistical quantities of this quality of fuel, and as you 11 saw, have produced test quality fuel, which is considerably 12 better than the specifications. The U.S. has not produced 13 this reference fuel in statistical quantities. It has, of 14 course, produced Fort St. Vrain, which is the basis of our 15 experience. The Germans have also used production in AVR.
16 That is their commercial offering at the same specification 17 as we are offering.
18 MR. SILADY: Thank you.
19 The one event I do want to go through is defined 20 in this fashion.
21 It begins with a moderate, that is, a very small 22 primary coolant leak. Moderate is a relative te rm , but this 23 is less than a square inch. The reactor is tripped, based 24 on high radioactivity in the reactor building or low 25 pressure in the primary circuit. The specifics of this ace. FEDERAL. REPORTEns, INC.
202 147 1700 Nationside Cmcrage N NLI)MM6
1 a
6850 12 11 144 DAVbw 1 event is that force core cooling is lost. You recall from 2 Neylan's presentation that we can remove the decay heat, not i 3 only with the heat transport system but with the shutdown i 4 cooling system, but in this particular event, both of those 1
l 5 are lost.
6 The decay heat that is removed via conduction and i
, 7 radiation to the reactor cavity cooling, and it's only with
- 8 conduction radiation and not convection in this particular f 9 event, because we are depressurizing, because of the leak, 4
10 circulating activity is then released, and we get an i
11 incremental release from the fuel, as the core temperatures 12 exceed normal operating levcis. -
l 13 Let me show you those temperatures.
! 14 DR. SIESS: Is this the worst case for Fort l 15 St. Vrain, no pressure, no circulation?
16 MR. SILADY: Yes; that's known as Design Basis 17 Accident No. 1 Fort St. Vrain.
1 18 l
l 19 l
20 l
1 i
21 1
22 L
23 24 25 Acti Fl!DliRAL RE! PORT!!RS, INC.
202 347 37m Nationw ale Cmcrage lun3M ud6
4 6850 13 01 145 DAVbur 1 (Slide.)
2 This diagram shows the traditional 3 characteristics of HTGR's slow time response. The chart 4 goes to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> and extends from the normal operating 5 range of the fuel, which on average is roughly 700 C but the 6 peak could be at 1100 C, and one sees that temperatures do 7 go up in order for the heat to be conducted and radiated out 8 through the side of the vessel. But at about 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> the 9 temperatures turn over and come down.
10 The average temperature is considerably lower 11 than the maximum. This is shown in the next graph --
12 (Slide.)
13 -- via an isotherm map on a cutaway and RZ 14 geometry.
15 L me orient you.
16 This is the axial direction. This is the radial 17 direction.
18 The active core, that which has the fuel, then is 19 within this annular region, which shows up here as a 20 rectangle.
21 The peak temperature occurs just on the inside 22 boundary near the graphite center column.
23 tiote that much of the core is considerably below 24 1600 C and since volumetrically this out here contributes t
A_-
s 25 much more to the average temperature, the average ACE-FEDERAL, REPORTERS, INC.
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6850 13 02 146 DAVbur 1 temperature is even lower than one would first see by 2 looking at this.
3 The side reflector continues to be lower in 4 temperature, and the normal gradient with the heat being 5 transferred through the vessel and out to the reactor cavity 6 cooling system.
7 (Slide.)
8 Repeating the viewgraph that Tony Neylan showed i 9 you, the fission gas release from particles is seen to be 10 negligible up to very high temperatures, when the silicon 11 carbide thermally begins to degrade. This is one component
~
12 of a release then from the core during this event. No
/ l
(,' 13 I incremental failure essentially.
{
14 Other components are the release from the i
15 f contamination or initially failed particles before the event l
16 l occurs. All of those contributions are taken into account.
I 17 (Slide.)
18 And the result over time is the following graph, 19 where now the time scale is still 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> and those are 20 cumulative Iodine-131 in curies. Iodine-131 contributes 21 about 80 percent of all the iodines and therefore 80 percent 22 of the thyroid dose. One sees a very slow release then.
23 And the top graph, the solid one is from the 24 fuel. Its peak value, which occurs hundreds of hours out f
25 into the event, is roughly 400 curies compared again to our ACE FEDERAL REPORTERS, INC.
202-347-3700 Nationwide Coserage 800-33M446
6850 13 03 147 DAVbur 1 long-term allowable or corresponding curies to the 150 rem.
2 One sees some margin there.
3 That, basically, is the approach then to removing 4 core heat and retaining the fission products --
S (Slide.)
6 -- within the fuel.
7 DR. CARBON: Is this for the case where you lose 8 all forced convection and you shut down?
9 MR. SILADY: Yes.
10 DR. CARBON: There are no active components?
i i
11 j MR. SILADY: None.
)
12 ! There are many other accidents -- and we could
- l
.i i
('
mz 13 t have a day's session on looking at the cases -- when it is l
14 pressurized or different variations. But I went to one that k
s 15 shows you the functions and is going to show you the design 16 l selections that were intentionally chosen in order to meet 8
17 our top level regulatory criteria.
18 If one looked at other events, one would find 19 that there are two other functions that are needed in order 20 to keep the fission products within the fuel. That is 21 control reactivity or heat generation and control chemical 22 attack.
23 And so by looking at a spectrum of events, 7 s 24 examining those events, one finds the functions, and one can
- 25 drive this down further. In term of perhaps for controlling ACE FEDERAL REPORTERS, INC.
202-347-3700 Nationwide Cos erage 8(n3346646
6850 13 04 148 DAVbur 1 reactivity, one needs to maintain core geometry, for 2 instance.
3 That is the kind of thing we are doing, and again 4 this is the basis for developing the principal design 5 criteria.
6 I would like to now move into the third and final 7 section, which is equipment classification. I would like to 8 start again with some definitions.
9 (Slide.)
10 Safety-related structures, systems, and 11 components are those performing these functions that we rely
- 12 upon in order to meet 10 CPR 100 for the design basis I
x,/ 13 events.
14 l So you see again this is tied to the other two l
15 parts, d
16 l The safety-related SSEs are described and their i
17 limiting design conditions; that is, the conditions under 18 which the safety-related equipment alone can keep you within 19 10 CFR 100, are evaluated in the PSID.
20 There are three steps in this third and final 21 part of the methodology.
22 (Slide.)
23 The first is one I will go into a little bit more
-- 24 than the other two, to start with each design basis event
(
, -) 25 and classify as safety related those systems, structures or ACE-FEDERAL REPORTERS, INC.
202-347-3700 Nationwide Coverage Sub336W>46
6850 13 05 + 149 DAVbur 1 components that are needed for compliance with the dose 2 criteria of 10 CPR 100.
3 (Slide.)
4 Returning to our risk plot and to the point for 5 which I showed you the transients, this particular event 6 here, CC-sub-P-dash-9. If it were not for the radionuclide 7 retention function of controlling the release, the point 8 would be out here and we would violate 10 CPR 100. So we l
9 saw that we had to remove core heat and keep the fission l
10 t products within the fuel particles.
11 gI So our design selections in this case, then, are 12 some form of alternate core cooling and the fuel quality (d 13"0 I
itself.
14 j Let me go into another layer of depth, then.
,1 15 g (Slide.)
0 You have already seen this in terms of the first 16 g 17f two, the specific numbers in that backup viewgraph of the I
18 design selection that we are specifying as safety related.
19 There is also a coating failure fraction; that 20 is, one can back out -- that would be allowable during 21 normal operation and transients similarly -- to keep within 22 those curies that corresponds to 10 CFR 100.
23 - The fuel quality, though, was not the only thinq l
,x 24 ' that had to be done.
I 25 (Slide.)
ACE-FEDERAL REPORTERS, INC.
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( ,),DAVbur 1 One also had to be able to remove the core heat.
2 These are the design selections, then, that were made that 3 would become safety related in order to do that.
4 The core power density is quite low, 5.9 watts 5 per cc. The annular geometry at this power level, 1.65 6 meters in inner diameter and 3.5 outer. Graphite, with its 7 heat capacity and its conductivity properties -- this should 8 read low alloy steel -- that is a vessel that the heat can 9 conduct out through. And a heat sink, which in this case we 10 have chosen to be the reactor cavity cooling.
11 I would like now to talk about why we chose this fg 12 decay heat removal system to be the one that is l
/\,_ -) 13 safety-related.
14 (Slide.)
15 You can imagine that for each of the functions 16 there are corresponding processes and examination of 17 events. I am taking just one slice through it as an 18 example.
i 19 But for the function to remove core heat, one 20 examines the spectrum of design basis events in which this l
21 function must operate in order to be within 10 CFR 100, and 22 one looks at all the different systems that are available to 23 perform that function.
l 24 So these are the system structures and components
(~
\s 25 that are available -- in this case for heat removal, mostly ACE-FEDERAL REPORTERS, INC.
! 202-347-3700 Nationwide Coserage NO-33MM6
6850 13 07 151 DAVbur 1 systems to remove decay heat in this particular first 2 event.
3 But I didn't go into the design basis event one.
4 All of these were available. Each of them alone could have 5 removed the decay heat.
6 So the "yes" is, yes, it is available to perform 7 the decay heat removal function. Not all of them are 8 I needed. Any one of them could have done it.
9 In design basis event five, it started off with l
10 l the loss of the main loop. So it wasn't available.
11 In design basis event eight there was only one of 12 these systems that could perform this function to the level
[~ )
t i
(' 13 $ required, and it was the reactor cavity cooling in the 0
14 Q passive mode.
il 15 y What one does then is look across and make a e
choice of whicn system or systems one makes safety related 16 l d
17 ; in order to satisfy this function, in order to satisfy the 18 top level regulatory criteria of 10 CPR 100 for design basis i
i cvents.
19 l I
20 This example is a kind of simplified. It just 21 works out that there is only one choice that covered the 22 whole spectrum of events. So the reactor cavity cooling was 23 i chosen as safety related.
-, 24 DR. SIESS: Do you only have one system, then?
_/ 25 MR. SILADY: The entire plant, which does all the ACE FEDERAL REPOR TERS, INC.
202-347-3700 Nationwide Cmerage 800 336-(M6
6850 13 08 152 DAVbur 1 functions and satisfies all the requirements of the user and 2 the regulatory body includes all these systems.
3 DR. SIESS: If I look at DBE-8, you then say one 4 system to remove decay heat is enough.
5 MR. SILADY: In each of these events, if this is 6 the only system you have, it is sufficient to keep the doses 7 within 10 CFR 100.
8 DR. SIESS: I am sorry. I am looking at DBE-8, t
9 where there is only one "yes" in the column. Now, I have 10 one system.
11 Where is the defense in depth?
s 12 MR. SILADY: The defense in depth is in the fact
- u. 13 that one has the fuel particles and the graphite core and 14 the other characteristics in order to provide backups to the i
15 reactor cavity cooling.
16 DR. SIESS: If that system fails, what does that 17 ! do to your point on your curve, on your plot?
l 18 I MR. SILADY: This is the system that we need.
19 DR. SIESS: And if it fails?
20 MR. SILADY: I believe the specifics of that 21 event are that you start off pressurized and you have a loss 22 of the main loop and you have a loss of the shutdown cooling 23 system.
7x 24 One then needs to have the reactor cavity cooling i
s_ 25 removing the energy in that case because the system is l
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50 13 09 ,
153 DAVbur 1 pressurized.
2 If one does not have that and if one does not 3 somehow repair whatever went wrong with the passive system, 4 the operator can intentionally depressurize the system, in 5 which case it is very similar to these other design basis 6 events, where even dissipation of heat to the ground is 7 sufficient.
8 DR. SIESS: So removing that capability moves you 9 down? {
I 10 MR. SILADY: Off the range.
11 DR. SIESS: Of the probability curve and 12 y potentially could move you over on the dose curve?
6 s 13 h MR. SILADY: That event doesn't show up in the 14 design basis region.
15 3 DR. SIESS: But it would push you into that other 16 l region in terms of probability?
E 17 l MR. SILADY: That is correct.
i 18 DR. SIESS: And of coursa, I assume that system 19 itself has redundant components?
20 MR. SILADY: Yes. You will see this afternoon 21 the level of redundancy. Yes.
22 DR. SIESS: Okay.
23 (Slide.)
,-s 24 MR. SILADY: The final two steps, just briefly.
\'
_/
25 One has to make sure that events are kept within ACE FEDERAL REPORTERS, INC.
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6850 13 10 154 8DAVbur 1 10 CFR 100 of the design basis events, and one specifies the 2 function and the design selection then as safety related.
3 Similarly, you wouldn't want to have an event 4 lower in frequency with doses greater than 10 CFR 100 that 5 could, if it were not made safety related, cause you to 6 exceed the 10 CFR 100. So one wants to keep that event out 7 of the design basis region.
So this step reads that for each emergency 8{
9' planning basis event with consequences greater than 10 CFR
(
10 100 classify as safety related those design selections that 4
11 are needed to assure that it is low in frequency, that it is f 12 below the design basis region.
, I
'j 13 ) The final step is to classify as safety related 14 !, ter each design selection classified as safety related to s
15 5 determine the limiting design conditions for its operation 16 ] by examining all its associated DBEs.
0 17 !! That gets back to your question. We would look i
18 at all those design basis events, the one that was 19 ! pressurized and the others that were depressurized, and 20 calculate the way in which the reactor cavity cooling system 21 has to respond. That becomes its limiting design 22 condition.
23 I would like to summarine then. I don't know if o 24 I have made any time, as I tried to do. But here is our i i l 25 summary.
ACE-FEDERAL REPORTERS, INC.
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,s, DAVbur 1 (Slide.)
2 The method uses the functional analysis of the 3 integrated approach, PRA, the risk assessment, and 4 reliability tools and the resulting design selections to 5 show compliance with the top level regulatory criteria. The 6 key is it always is linked back into that top level 7 regulatory criteria.
8 The method provides a systematic, traceable
! 9 process to derive the licensing bases specific to the 10 modular HTGR.
4 11 The application of the method demonstrates that 4
pg 12 the modular HTGR is emphasizing retention within-the fuel
- (\m- 13 with passive features.
I 14 And the approach is consistent with the draft i 15 advanced reactor policy.
16 Any additional questions?
17 DR. SIESS: I am still thinking in terms of your 18 frequency consequence diagrams. I know there is always room 19 for argument and uncertainty about reliability and of the 20 probability curve, and I guess until we got into this light 1
21 water reactor source term I thought there was less argument I 22 about what the consequences might be, and I still have some 23 hope that for the light gas cooled reactors there might be 24 less argument.
N 25 Can you conceive of a maximum release without l
ACE-FEDERAL REPORTERS, INC.
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1
6850 13 12 156 DAVbur 1 assigning a probability to it?
2 MR. SILADY: In essence, that is what we are 3 doing here. We are trying to come up with a scheme that 4 guarantees that the doses are below the protective action 5 guidelines with a design that performs those functions for a 6 wide spectrum of events, going all the way down to 5 times 7 10 to the minus 7 th.
8 DR. SIESS: Right now it seems to me that if you 9 scram the reactor and you can't dump the stuff into the 10 earth without a circulating system, you can't get too much 4
11 l out of there in the way of fission products.
g c ,
12 MR. SILADY: That is correct. The design is set
/ a "N / 13 ) up so that it can remove the heat passively without h
14 0 releasing from the fuel particles.
O h
1 5 ll DR. SIESS: To ge t more than that out you would N
have to have a failure to scram?
16 l l
17 i MR. SILADY: I don' t even know if that is that 18 h crucial.
i 19 ; We have taken a look at the negative temperature 20 coefficient of the system and looked at the event I have 21 just showed you plus others without trip, and the 22 temperatures go up because it takes a few minutes in order 23 for the negative temperature coefficient to turn the x 24 transient around. But they only go up by 50 degrees, or
'_- 25 something.
ACE-FEDERAL REPORTERS, INC, 202-347-3700 Nationside Coserage 800-336-St6
c850 13 13 157 t i
,/ DAVbur 1 So essentially our releases are the same.
2 I didn't get off into those events and get off 3 into controlling heat generation today because I was really 4 trying to get the feedback on the approach.
5L But in another session we could go into the high l
6l degree of passivity that is in the design for controlling 7 heat generation.
8l DR. SIESS: Let me say that I think you have I
9 got -- the approach is quite transparent -- questions about 10 l it.
I 11 I understand it. I think it is a very helpful 12 l way of looking at it, and by being transparent that simply l
it 13 l opens you to a lot more questions.
14 f (Laughter.)
l i
15 g DR. SIESS: But I think that is an object of the 4
16 l game.
17 MR. SILADY: Thank you very much.
18 ' (Whereupon, at 12:55 p.m., the subcommittee was 19 recessed, to go into closed session.)
20 21 22 23
,s 24
)
[ j f 25 ACE-FEDERAL REPORTERS, INC.
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CERTIFICATE OF OFFICIAL REPORTER
,/ \
i 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 SUBCOMMITTEE ON ADVANCED REACTORS DOCKET NO.:
PLACE: Washington, D. C.
DATE: Thursday, January 30, 1986 were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission.
(sigt) i ef (TYPED)
DAVID L. HOFFMAN Official Reporter ACE-FEDERAI BEPORTERS, INC.
Reporter's Alf111ation i
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l i LICENSING BASIS EVENT SELECTION l
l STEP 1:
i DEFINE THREE REGIONS ON A FREQUENCY-CONSEQUENCE RISK PLOT STEP 2:
COMPARE RISK ASSESSMENT OF THE DESIGN TO THE FREQUENCY CONSEQUENCE RISK PLOT
( J FSVIEW.VWG S 23-JAN-86
i F 3 EVENT DEFINITIONS (CONTINUED) i EMERGENCY PLANNING BASIS EVENTS (EPBEs)
EVENTS OF LOWER FREQUENCY THAN DBEs, NOT EXPECTED TO OCCUR IN THE UFETIME OF ALL STANDARD HTGRs CONSEQUENCES REALISTICALLY ANALYZED IN PRA AND CHAPTER 13 OF SARs (PSID) FOR COMPUANCE WITH PAG DOSE UMITS
< - - - J rs .v. . 22- -
~
O O O F 3 EVENT DEFINITIONS (CONTINUED)
DESIGN BASIS EVENTS (DBEs)
EVENTS OF LOWER FREQUENCY THAN AOOs, NOT EXPECTED TO OCCUR IN THE LIFETIME OF THE PLANT CONSEQUENCES CONSERVATIVELY ANALYZED IN CHAPTER 15 OF l SARs (PSID) FOR COMPLIANCE WITH 10CFR100 FREDS.VWG 3 20-JAN-86
" 3 l
PRESENTATION OUTLINE
- LICENSING BASIS EVENTS
-DEFINITIONS
-METHOD / APPLICATION e REQUIRED FUNCTIONS
-DEFINITIONS
-METHOD
-APPLICATION l e EQUIPMENT CLASSIFICATION
-DEFINITIONS
-METHOD / APPLICATION e
SUMMARY
l L J g g ....._3
,29
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I DEVELOPMENT OF LIGENSING BASES PRESENTED TO THE ACRS JANUARY 30,1986 FRED A. SILADY GA TECHNOLOGIES, INC.
L J FGRAPH,VWG 1 24 -J AN -86
,, - - - , , - - - , , - - -,- , - - - - - - - - - - m
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EVENT DEFINITIONS (CONTINUED)
ANTICIPATED OPERATIONAL OCCURRENCES ( AOOs )
EVENTS EXPECTED ONCE OR MORE IN THE PLANT LIFETIME CONSEQUENCES REALISTICALLY ANALYZED IN CHAPTER 11 OF SARs (PSID) FOR COMPLIANCE WITH 10CFR50 APPENDIX l e J g g mos.m 2 20_;g
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LICENSING BASIS EVENT DEFINITIONS l LICENSING BASIS EVENTS (LBEs)
THE OFF-NORMAL OR ACCIDENT EVENTS USED FOR DEMONSTRATING DESIGN COMPLIANCE WITH THE TOP-LEVEL REGULATORY CRITERIA COLLECTIVELY LBEs ARE ANALYZED IN PRAs FOR DEMONSTRATING
, COMPLIANCE WITH INTERIM SAFETY RISK GOALS LBEs ENCOMPASS THE FOLLOWING THREE EVENT CATAGORIES l
1 FSVIEW.VWG 1 24-J A N-86
l LICENSING APPROACH.
SUMMARY
THREE STEPS:
i
- DONE.
- 2) DEVELOP PROCESS TO DERIVE LICENSING BASES SPECIFIC TO THE MHTGR WHICH ENSURE THAT TOP-LEVEL CRITERIA ARE MET.
- DONE.
- 3) APPLY PROCESS TO IDENTIFY MHTGR LICENSING BASES.
- IN PROGRESS.
F 3
. .. _ ~ .. - .
LICENSING METHODOLOGY
SUMMARY
e USE PRA TO IDENTIFY LIKELlHOOD 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 e CHOOSE DESIGN SELECTIONS TO ACCOMPUSH REQUIRED FUNCTIONS TO MEET THE TOP LEVEL REGULATORY CRITERIA
( )
a m ..- 2 2,-n
HTGR DESIGN & LICENSING APPROACH 1
USER TOP-LEVEL REQUIREMENTS REGULATORY l CRITERI A l LICENSING BASIS f , , d INTEGR ATED APPRO ACH 4 l BRIDGE 4-+
1 LICENSING B ASIS EVENTS l
EQUIPMENT CLASS 1
l ENGINEERING PRODUCT OTHER BASES l PLANT DESIGN, ETC.
l l
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C) O O l F 3 I @
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1 OVERVIEW OF THE LICENSING BASES DEVELOPMENT I
l PRESENTED TO THE ACRS i
l JANUARY 30,1986 i
j FRED A. SILADY GA TECHNOLOGIES, INC.
i N Y
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AAO SELECTION / EVALUATION i
STEP 3:
IDENTIFY AS AAOs THOSE FAMlUES OF EVENTS WITHiN THE AAO REGION STEP 4: ,
EV LUATE THE CONSEQUENCES OF THE SELECTED AOOs REALISTICALLY e
FRED 2.VWG S 24-J AN-8 6 4 -
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O O IDENTIFY PC-6 AS ma 1 %,,a ANTICIPATED OPERATIONAL OCCURRENCE 10 '
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F 3 DESIGN BASIS EVENT SELECTION / EVALUATION l STEP 5:
IDENTIFY AS DBEs THOSE FAMILIES OF EVENTS WITHIN THE DESIGN BASIS REGION STEP 6:
IDENTIFY AS DBEs THOSE EVENTS WITH AGREED UPPER OR LOWER BOUND FREQUENCIES THAT LIE WITHIN THE DESIGN BASIS REGION AND SATISFY STEP 5 STEP 7:
EVALUATE THE CONSEQUENCES OF THE SELECTED DBEs CONSERVATIVELY L J FRED 2.VWG 6 2 3 -J A N -
O O O f 3
@~..-
EXAMPLE SELECTION OF DESIGN BASIS EVENT DEPRESSURIZED CONDUCTION COOLDOWN WITH RCCS CCp -9
! EVENT SEQUENCE:
i
- 1. MODERATE PRIMARY COOLANT LEAK OCCURS
- 2. REACTOR IS TRIPPED
- 3. FORCED CORE COOLING IS LOST ( BOTH HTS AND SCS )
(
- 4. DECAY HEAT REMOVED VIA CONDUCTION AND RADIATION TO REACTOR CAVITY COOLING
- 5. CIRCULATING ACTIVITY IS RELEASED
- 6. INCREMENTAL RELEASE FROM FUEL AS CORE TEMPERATURES EXCEED NORMAL OPERATING LEVELS
( J CC9.V WG 2 23-JAN-86
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EVALUATE THE CONSEQUENCES OF THE SELECTED EPBEs REAUSTICALLY
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TO LATENT CANCER RISK GOAL REGION EVENT EVENTS PER CANCERS RISK PLANT YEAR PER EVENT A00 PC-10 1.1 5X10-' 6X10-'
A00 PC-6 0.17 7X10-8 1.2X10-a 10CFR100 PC-4 1X10-2 IX10-7 1X10-'
10CFR100 PC-9 2X10-3 2X10-s 4X10-
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F REQUIRED FUNCTIONS DEFINITION THOSE FUNCTIONS NEEDED TO LIMIT RADIONUCLIDE RETENTION TO MEET 10CFR100 DOSES FOR DBEs BASIS FOR PRINCIPAL DESIGN CRITERIA EVALUATED IN CHAPTER 3 0F SARs (PSID) FOR COMPLIANCE WITH TOP LEVEL REGULATORY CRITERIA
.... , mm
O r 3 SELECTION OF REQUIRED FUNCTIONS S T E P 1:
IDENTIFY THOSE RADIONUCLIDE RETENTION FUNCTIONS REQUIRED TO MEET TOP LEVEL REGULATORY CRITERIA AND USER / UTILITY SAFETY REQUIREMENTS STEP 2:
SELECT THOSE RADIONUCLIDE RETENTION FUNCTIONS REQUIRED TO MEET 10CFR100 DOSES FOR DBEs I
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f D EXAMPLE DBE EXAMINATION FOR REQUIRED FUNCTIONS DEPRESSURIZED CONDUCTION COOLDOWN WITH RCCS CCg -9 EVENT SEQUENCE:
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- 6. INCREMENTAL RELEASE FROM FUEL AS CORE TEMPERATURES EXCEED NORMAL OPERATING LEVELS
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) SAFETY RELATED STRUCTURES, SYSTEMS,AND COMPONENTS (SSCs)
- SAFETY RELATED SSCs ARE DESCRIBED AND THEIR LIMITING
) DESIGN CONDITIONS EVALUATED IN THE SARs (PS!D) l I
, -..-,...4
O O O F 3 SELECTION OF SAFETY-RELATED SYSTEMS STRUCTURES AND COMPONENTS STEP t FOR EACH DBE, CLASSIFY AS SAFETY-RELATED THOSE
, SSCs, NEEDED FOR COMPLIANCE WITH 10CFR100 DOSE l CRITERIA l
i ( )
FRED 2.VWG 4 27-J AN-86 4
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O O O l r 3 SELECTION OF SA:r-i: 1Y-RELATED SYSTEMS STRUCTURES AND COMPONENTS i
STEP 2:
! FOR EACH EPBE WITH CONSEQUENCES GREATER THAN i
10CFR100 CLASSIFY AS SAFETY-RELATED THOSE SSC
! DESIGN SELECTIONS NEEDED TO ASSURE THAT THE EVENT FREQUENCY IS BELOW THE DESIGN BASIS REGION STEP 3: ,
j FOR EACH SSC CLASSIFIED AS SAFETY RELATED, DETERMINE THE UMITING DESIGN CONDITIONS FOR IT'S OPERATION BY EXAMINING ALL IT'S ASSOCIATED DBEs L J
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i 1 SAFt-1Y RELATED SSCs NEEDED FOR RADIONUCLIDE l RETENTION WITHIN FUEL i
FUEL QUAUTY AS-MANUFACTURED:
! CONTAMINATION FRACTION i
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- 1.65 M 10 i 3.50 M OD l
l CORE HIGH TEMPERATURE GRAPHITE l MATERIAL PROPERTIES VESSEL MATERIAL PROPERTIES CARBON HEAT SINK REACTOR CAVITY COOLING SYSTEM FGR.VWG 2 24-J AN -86
(
mm= % .mm REPRESENTATIVE CLASSIFICATION OF AN SSC AS SAFETY-RELATED FOR THE MHTGR RE0uiRED FUNCTION: REMDVE CORE HEAT 1 SSC AVAILABLE TO PERFORM DBE-1 DBE-5 DBE-8 DBE-11 DDE-14 FUNCTION ? PC-4 PC-9 CCL-N1 CCP-9 PC-5 CLASS.
- 1. MAIN LOOP YES YES
- 2. SHUTDOAN YES YES YES COOLING SYSTEM
- 3. REACTOR YES YES YES YES CAVITY COOLING l (ACTIVE MODE) i 4. REACTOR YES YES YES YES YES X CAVITY COOLIHG (PASSIVE MDDE) i 5. REACTOR YES YES YES YES j CAVITY &
. SURROUNDINGS
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FREDS.VWG 1 27-J AN- 86
- 4
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LICENSING BASIS
SUMMARY
l I e METHOD USES FUNCTIONAL ANALYSIS (IA), PRA AND l RESULTING DESIGN SELECTIONS TO SHOW COMPLIANCE WITH TOP LEVEL REGULATION CRITERIA i e METHOD PROVIDES SYSTEMATIC, TRACEABLE PROCESS
) TO DERIVE LICENSING BASES SPECIFIC TO MHTGR e APPLICATION OF METHOD DEMONSTRATES MHTGR EMPHASIS f
i ON RADIATION RETENTION WITHIN FUEL WITH PASSIVE
) FEATURES e APPROACH IS CONSISTENT WITH DRAFT ADVANCED REACTOR POUCY l
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i NUCLEAR STEAM SUPPLY SYSTEM (NSSS) l DESIGN OVERVIEW l
1 A. J. NEYLAN l DIVISION DIRECTOR :
1 i 1
l
! GA TECHNOLOGIES i -
i N Y l VWC1.VWG 1 23-JAN-86
f 3 MAJOR NSSS SYSTEMS
- REACTOR SYSTEM e HEAT TRANSPORT SYSTEM e VESSEL SYSTEM e SHUTDOWN COOLING SYSTEM ,
o REACTOR CAVITY COOLING SYSTEM (IN BOP) e PLANT CONTROL & PROTECTION SYSTEMS e FUEL HANDLING SYSTEM e REACTOR SERVICE SYSTEMS o V J
- vwci.vwo 4 23-uw f.
-o O O i
NUCLEAR STEAM SUPPLY SYSTEM CONFIGURATION DESIGN PARAMETERS V J VWG.YWG 3 20-JAN-86
f D M oA TN= m l CONFIGURATION l
SEPARATE REACTOR & STEAM GENERATOR STEEL VESSELS CONCENTRIC CROSS DUCT PRISMATlC FUEL - DOWNFLON ANNULAR CORE CORE BARREL SUPPORT STRUCTURE STEAM GENERATOR - ONCE THRU/ UPHILL BOILING TOP-MOUNTED MAIN CIRCULATOR - IN COLD LEG OF LOOP ELECTRIC MOTOR DRIVE MAIN CIRCULATOR SEPARATE SHUTDOAN COOLING SYSTEM - HEAT EXCHANGER AND CIRCULATOR BELON CORE LOCATED IN SILO IN BOP-NUCLEAR ISLAND 1
s y q .- .n q
_ _ _ _ _ _ _ _ _ _ _ . _ - _ _ - - - _ - - - - - - - - ~ ' ' - ~ - - - - - -
\
! y' o O
- CONTROL R00 DRIVE /
- REFUELINO PENETRATIONS l
\
f<
STEEL REACTOR IRCULATOR ACT0R CORE-350 MWs,t,?
! su.
i MODULAR HTGR sRur00wN GENERATOR HEAT EXCHANGER- i==mi , vEsstt
! PLANT j
g su, OUTLET C RCULA OR y STEA 99 y FEEDWATER
< .20 5 INLET
O O O ITI ITI
[n n\
@ lllll lllll 350 MWLt? MODULAR t i i HTGR PLANT HEllUM FLOW I
(^'
NORMAL OPERATION LA
-w 3 C3 t.
/ i-f e , ,
t H-303(1) 6-28-85 i
l NSSS HEAT BALANCE DIAGRAM 4 x 350 MWT MODULAR HTGR PLANT m GATechnologies M FROM OTHER NSSS BOP SG LOOPS y
~
1266 F 1005 F l STEAM.
916 PSIA 2515 PSIA l TURBINE 1,089,700 LB/H GENERATOR PSIA 1,248,100 LBlH ,1 REACTOR
s CORE STEAM GENERATOR : 352.15 MWT N
hf ( ) CONDENSER 350 MWT 491 F II
\ REJECT 912 PSIA HEAT i
380 F h PSIA FEE 0 WATER
! 9 PSIA PUMP i
" u TO 25 PSIA .2 PS G0 S 3.60 MWE
)
NSSS HEAT LOSSES 0.73 MWT LOSS TO ENVIRONMENT 0.25 MWT LOSS TO SCHE ClR U ATOR j H 491(3) 12-12-85 LL
- 9
i o O O F D 1
amo - .ma DESIGN PARAMETERS
! o CORE PONER RAT!NG (4 MDLS) 1400 MN(t'!
l o CORE PONER DENSITY 5.91W/ad
)
o HELIUM TEMP. CORE INLET 497 F o HELIUM TEMP. CORE OUTLET 1268 F 1
o HELIUM PRESSURE 925 psia o ClRCULATOR SPEED /HP 5,720 rpT)/4,800 o FEEDNATER PRESSURE 3,000 psia o FEED #ATER TEMP. 380 F o STEAM TEMP. 1005 F o STEAM PRESSURE 2515 psig o STEAM GENERATOR OUTPUT 352 MN( t )
L J VWGl.VWG 5 23-JAN-86
350 MWLt? MODULAR
_ , , _ _ REACTOR CORE CROSS SECTION CENTRAL REFLECTOR SIDE REFLECTOR REACTOR VESSEL
\N [/
o o ANNULAR
.o o .
ACTIVE CORE CORE BARREL o, o o 4-lt
( o oh"7.w'o -
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CONTROL ROD o: o o eff. -
-o CHANNELS
- (6 INNER o~ 24 OUTER)
SEISMIC KEYS- \ o o O o RSS CHANNELS (12)
H NNE S Y BORONATED PINS
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! HEAT TRANSPORT SYSTEM (HTS) l SHUTDOWN COOLING SYSTEM (SCS)
- REACTOR CAVITY COOLING SYSTEM (RCCS) l g .
VWG.VWG 2 20-JAN-86
l 4
I!!!I I!!!I l DECAY HEAT t i REMOVAL WITH SHUTDOWN COOLING SYSTEM if;.
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DECAY HEAT REMOVAL. g BY CONDUCTION /
AND RADIATION
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L Is 85 _
CORE TEMPERATURES REMAIN BELOW
@" DESIGN LIMITS DURING ACCIDENTS 3000 1600 -
WITH RADIATION &
1400 CONDUCTION TO RCCS -
2500 g *, 1200 -
O h 33 -
2000 y
!!s =
gg 1000 -
.g 800 -
WITH SHUTDOWN COOLING SYSTEM OPERATING (FOR DEPRESSURIZED C00LDOWN)
~
L i I I I 1000 0 200 400 600 800 1000 TIME PAST REACTOR TRIP, HR ,
H-323(1) i g 12-6-85
~
@ e e
l o O O C0ATED FUEL PARTICLES MAINTAIN INTEGRITY UP TO 18000C l
1.0 P TEST RESULTS FOR g
! TRISO COATED UC0 0.8 -
FUEL PARTICLES m
g 0.6 -
0 I E
! E o j g 0.4 -
i
<c l NORMAL l 0.2 - PEAK FUEL g
. OPERATING
! TEMPERATURE g e I ' ' '
0 1000 1200 1400 1600 1800 2000 2200 2400 2600 TEMPERATURE (8C )
l N-150(9) i 9-4-84 i
i j
i
! ACRS BRIEFING OBJECTIVES i
e BRIEF ACRS ON APPROACH TO DESIGN l e BRIEF ACRS ON LICENSING APPROACH AND METHODOLOGY WHICH HAS BEEN PROPOSED TO NRC STAFF .
e BRIEF ACRS ON MHTGR DESIGN STATUS
! e RECEIVE ACRS COMMENTS ON PROPOSED LICENSING APPROACH AND METHODOLOGY
O O O 2
PROGRAM OBJECTIVE ,
DEVELOP HTGR's FOR BROAD RANGE OF APPLICATIONS IN SUPPORT OF
- COMMERCIAL / USER INTERESTS IN SAFETY t l AND HIGHER TEMPERATURE CHARACTERISTICS OF THESE PLANTS.
O O O HTGR PROGRAM PARTICIPANTS
- GA TECHNOLOGIES INC.
- GENERAL ELECTRIC COMPANY e COMBUSTION ENGINEERING, INC. ,
e STONE & WEBSTER ENGINEERING, INC.
e BECHTEL GROUP, INC.
e OAK RIDGE NATIONAL LABORATORY e IDAHO NATIONAL ENGINEERING LABORATORY e EG&G IDAHO, INC.
- GAS-COOLED REACTOR ASSOCIATES
o a gJ t
I 'ff i '
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p CONTROL ROD DRIVE /
REFUELING PENETRATIONS l .I
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IRCULATOR ANNULAR
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MODULAR F.. .
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q' i* Ki VESSEL SHUTDOWN HEAT EXCHANGER A 0 /..v -
CROSS DUCT SHUTDOWN g~j CIRCULATOR Y '
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- ._)
FEEDWATER INLET l
9 O O NRC INTERACTIONS I I FY 1985 l FY 1986 l FY 1987 LICENSING
! PLAN
""SE'oS$^' '
I I I A G R E E !4 E N T " - AGRE MENT ON ON T0 3-LEVEL LICEN-ING BASES TECHNICAL -
APPROACH l l 1 l l l PSID DESIGN / TECHNOLOGY l
\7 FAMILIARIZATION ,
I l l l LICENSABILITY j
STATEMENT l l g PSER 7 i
DESIGN / TECHNOLOGY . a >
l REVIEW
- I I i
l--. _ - _ - - - _ _
em i # 0 O
PROCEDURAL APPROACH INTERACTION LICENSING PLAN DRAFT COMPLETE l
l NRC ADVANCED REACTOR .
COMPLETE POLICY ISSUED FOR COMMENT 4
l LICENSING PLAN FORMAL SUBMITTAL COMPLETE 1
1 INDUSTRY COMMENTS ON POLICY COMPLETE NRC ACCEPTANCE OF LICENSING COMPLETE PLAN
~~
e
~
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O O DESIGN / TECHNOLOGY FAMILIARIZATION l lSSUE BRIEFING MODULAR HTGR DESIGN 12/85 l FUEL 3/86 DECAY HEAT REMOVAL 5/86 REACTIVITY CONTROL 5/86 i CORE SUPPORT STRUCTURE 5/86 j ISI 7/86 i
WATER / AIR INGRESS 7/86 CONTAINMENT / CONFINEMENT 8/86 -
BOP CLASSlFICATION 8/86 MULTIPLE MODULE CONTROL 8/86 STANDARD PLANT ISSUES 8/86 ACRS BRIEFING 9/86
O O O l
DESIGN & TECHNOLOGY FAMILIARIZATION (conti DOCUMENT DATE l TECHNOLOGY PLAN .
9/86 PRA 9/86 PSID
- OUTLINE COMPLETE
- FULL SUBMITTAL , 9/86 DESIGN & TECHNOLOGY REVIEW PSER 6/87 LICENSABILITY STATEMENT 9/87
a O O O l HTGR DESIGN & LICENSING APPROACH j
l USER TOP-LEVEL REGULATORY REQUIREMENTS CRITERI A -
f
. LICENSING BASIS PRINCIP AL DESIGN CRITERI A INTEGR ATED APPRO ACH 4 l BRIDGE (-+
- l
- LICENSING B ASIS EVENTS
! , , EQUIPMENT CLASS I i l ENGINEERING PRODUCT OTHER B ASES l
PL ANT DESIGN, ETC.
i
-.r- . . _ _ . _
SAFETY PHILOSOPHY e PROVIDE DEFENSE-IN-DEPTH THROUGH PURSUIT OF 4 GOALS:
1 - MAINTAIN SAFE PLANT OPERATION ,
l 2 - MAINTAIN PLANT PROTECTION 3 - MAINTAIN CONTROL OF RADIONUCLIDE RELEASE 4 - MAINTAIN EMERGENCY PREPAREDNESS e GOAL 1 TO BE ACHIEVED BY HIGHLY REllABLE j;
OPERATION AND WITH WELL TRAINED PERSONNEL i e GOALS 2 & 3 TO BE ACHIEVED THROUGH i
UTILIZATION OF INHERENT CHARACTERISTICS AND
! PASSIVE SAFETY FEATURES I
- e GOALS 1 - 3 TO BE ACHIEVED SO WELL THAT'
! MINIMAL REllANCE NEED BE PLACED ON GOAL 4 i
i O O . O i
LICENSING APPROACH
SUMMARY
THREE STEPS:
l - DONE.
- 2) DEVELOP PROCESS TO DERIVE LICENSING BASES SPECIFIC TO THE MHTGR WHICH ENSURE THAT '
TOP-LEVEL CRITERIA ARE MET.,
- DONE.
- 3) APPLY PROCESS TO IDENTIFY MHTGR LICENSING
! BASES.
- IN PROGRESS.
I ._
O O 9 PURPOSE OF Tile INTEGRATED APPROACll THE INTEGRATED APPROACH IS USED T0:
o DEVELOP REQUIREMENTS o EVALUATE DESIGNS SELECTED TO MEET REQUIREMENTS o COMMUNICATE l
O O 9 SAVINGS FROM THE INTEGRATED APPROACH THE SAVINGS ENVISIONED FROM THE USE OF THE INTEGRATED APPROACH ARE DUE TO:
t e A CLEAR UNDERSTANDING BY THE DESIGNERS, CONTRACTORS AND OPERATORS OF WHAT THEIR ROLES AND RESPONSIBILITIES ARE.
- AN EARLY IDENTIFICATION OF INTERFACES WHICH REDUCES THE RISK OF LATER MORE COSTLY REVISIONS.
I e VISIBILITY OF THE BASIS FOR DESIGN REQUIREMENTS.
e EllMINATION OF UNJUSTIFIABLE RETROFITS.
e JUSTlFICATION FOR, OR DELETION OF, DEVELOPMENT G-817(6) 8-13-84
O O 9 4
PLANT GOALS 4
e i
SAFE ECONOMICAL POWER MAINTAIN MAINTAIN MAINTAIN MAINTAl# EMERGENCY CONTROL OF PLANT PLANT RADIONUCLIOE PREPAREDNESS PROTECTION OPERATION RELEASE GOAL 4 GOAL 3 GOAL 2 GOAL 1 G-8171161 8-13 84
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- DEVELOP FUNCTIONS AND REQUIREMENTS AND MAKE DESIGN j SELECTIONS TO MEET GOAL 1 REQUIREMENTS TO i
MAINTAIN NORMAL PLANT OPERATION.
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- e DEVELOP GOAL 2 FUNCTIONS AND REQUIREMENTS TO MAINTAIN PLANT PROTECTION
- DETERMINE IF GOAL 1 DESIGN SELECTIONS MEET THE GOAL 2 FUNCTIONS AND REQUIREMENTS e IF REQUIRED MODIFY OR SUPPLEMENT GOAL 1 DESIGN i SELECTIONS D R t002.VWC 2 27-u AT-85
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O O O HTGR BRIEFING FOR THE ACRS SUBCOMMITTEE ON ADVANCED REACTORS JANUARY 30,1986 TOP-LEVEL REQUIREMENTS / CRITERIA FOR HTGRs PRESENTATION BY
- ARCHIE P. KELLEY, JR.
GAS-COOLED REACTOR ASSOCIATES
I i
TOP-LEVEL REGUIREMENTS/ CRITERIA FOR HTGRs ePURPOSE j e UTILITY / USER REQU.lREMENTS I
e REGULATORY CRITERIA .
1 l - BASIS FOR SELECTION
- PROPOSED CRITERIA ,
i l
l
TOP-LEVEL USER CRITERIA OVERALL - SAFE, ECONOMICAL NUCLEAR POWER e 10% ECONOMIC ADVANTAGE OVER COAL ALTERNATIVE e SITING ENVELOPE COVERING 85%
OF U.S. SITES e SERVICE LIFE OF 40 YEARS ,
GOAL 1 - MAINTAIN SAFE PLANT OPERATION EQUIVALENT UNAVAILABILITY OWING TO PLANNED OUTAGES LESS THAN 10%
p - _ . , m-
i l 0 0 0-i i
TOP-LEVEL USER CRITERIA { cont.)
GOAL 2 - MAINTAIN PLANT PROTECTION ,
i e EQUIVALENT UNAVAILABILITY OWING TO UNPLANNED OUTAGES LESS THAN 10%
e ANNUAL EXPECTED VALUE OF DAMAGE LESS THAN INSURANCE PREMlUM OF $4.5 MILLION e MEAN LIKELIHOOD OF A LOSS OF A SINGLE j REACTOR LESS THAN 10-5 PER YEAR ,
I l
O -
o 'o TOP-LEVEL USER CRITERIA (cont.)
l GOAL 3 - MAINTAIN CONTROL OF RADIONUCLIDE RELEASE l
MEET TOP-LEVEL REGULATORY CRITERIA WITHOUT CREDIT FOR SHELTERING OR EVACUATION OF PUBLIC CURRENTLY INTERPRETED AS REQUIRING THAT PAG DOSES BE MET FOR EVENTS WITH MEAN FREQUENCIES l GREATER THAN 5 x 10-7 PER YEAR l
l o
w- -.- . ,
O O O.
1 J
PROPOSED BASES FOR TOP-LEVEL CRITERIA SELECTION
- 2) CRITERIA MUST BE INDEPENDENT OF PLANT DESIGN
- 3) CRITERIA MUST BE QUANTIFIABLE
~
l
l O O O l PROPOSED SOURCES AND CANDIDATES i FOR7DP-LEVEE RE~GULATORY CRITERIA OVERALL - SAFE. ECONOMICAL POWER NUREG-0880: ,
! - INDIVIDUAL & SOCIETAL MORTALITY RISKS t
l - COST BENEFIT INVOKED ONLY IF MORTALITY
! RISK CRITERIA NOT MET -
i i t '
1
! O O O.
l PROPOSED S FOR TOPTEVEE_OURCES AND CANDIDATES REGULATORTCRITERIA(cont.)
GOAL 1 - MAINTAIN SAFE PLANT OPERATION e 10CFR20:
i
- PERMISSIBLE DOSE LEVELS & ACTIVITY CONCENTRATIONS IN UNRESTRICTED AREAS i
l e 10CFR50 APPENDIX 1:
I j - NUMERICAL DOSE GUIDELINES GOAL 2 . MAINTAIN PLANT PROTECTION e TO BE COVERED BY OCCUPATIONAL i
i EXPOSURE CRITERIA
~
i l !
i - - - .
PROPOSED SOURCES AND CANDIDATES l
FOIFTOP EEVEllEGUE~ATORTCRITERIA(cont.)
GOAL 3 - MAINTAIN CONTROL OF RADIONUCLIDE RELEASE i e 10CFR50 APPENDIX 1:
- APPLIED ON AN EXPECTED VALUE BASIS TO EVENTS ANTICIPATED TO OCCUR IN PLANT LIFETIME e 10CFR100:
- NUMERICAL DOSE GUIDELINES GOAL 4 - MAINTAIN EMERGENCY PREPAREDNESS i .
e EPA-520:
- PAG DOSES
~
e
_ _ a --____-----
NRR STAFF PRESENTATION TO THE O ACRS
SUBJECT:
STATLE OF INTERACTIGIS ON TE DOE ADVANCED HIGR PROGRA1.
DATE: JNitMRY 30,1986 PRESENTER: THU%S L. Klf6 O
PRESENTER'S TITLE / BRANCH /DIV: SECTIQ1LEAER SAETY PROGP#1 EVALLMTIG1 BRN10i,' DIVISIQ10F SAFETY REVIEW AND OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-7347 SUBCOMMITTEE: ADV# ICED [ ACTORS O
.a i
O t O'ERALL REVIEW PLAN REVIElf C0f:CEFTL'Il DESIGN AND KEY ISSUES OVER THE NEXT WO YEARS:
AGREE ON CRITERIA i
/.SSESS POTENTIAL OF THE DESIGN FOR SATISFYING THESE CRITERIA ASSESS PID PP0 GRAMS SUPPORTING THE DESIGN O -
ISSUE sea AND ticeNS^EiLitv STATEnerrt BASED ON THE REVIEW OF THE CONCEPTUAL DESIGN IDENTIFY ADDITIOM.L STEPS NRC MUST TAKE (INCLUDIb'G PESEARCH) TO BE READY TO PROCESS AN HTGR APPLICATION.
2 O
. - , .- ._ ,_,- ___,.-.-.-_--_- ..----__,_--- _ --_.--,-__.___= _ __. - _-.-, _ _ __- _-- _ __
i
- O m R m eRACn 0N SCHEDute ERIEFING TO SLEMITTAL NRC ACPS ITEM NRC STAFF TO NRC ACTION BRIEFIFE LICENSING APPROACH LICENSING PLAM - REV, 2 CTPLETE 2/86 N/A N/A TOP LEVEL CRITERIA C&FLETE COMPLETE 2/86 1/86 ERIDGING PETHODS COPPLETE 2/E6 3/06 1/86 ACCIDEis SELECTION CRITERIA C&FLETE 2/E6 3/86 1/86 LIE'S, SAFETY SSC'S
- ET10D REVIEW COMPLETE 2/86 3/06 1/f6
- REVIEW OF LBEs/SSCs 2/86 4/F6 N/A IT.JOR ISSUES DLCAY HEAT REMOVAL 2/86 TO BE N/A
! CCRE SL'PPORT STRUCTURE 3/86 COVERED fl/A L FEACTIVITY C0fGROL 3/86 IN PSID, il/A FL'EL U/86 PRA,8 N/A l ISI 4/E6 TECHNOLOGY N/A NJLTIPLE MODULE CONTROL 5/E6 PLAN N/A WATER / AIR INGRESS 7/85 N/A STANDARD PLANT ISSUES 8/86
N/A i C0lGAINFElR/CONFINEENT 8/86 9/EE i ECP CLASSIFICATION 8/86 9/E6 f
I
!O
.i r
l O ITEM BRIEFING TO StBilTTAL NRC ACRS NRC STAFF T0fEC ACTION ERIEFING ,
4 l CESIGN & TECM' OLOGY REVIEW
.I
!- BRIEFING ON REFERENCE CONCEPT C0PRETE -
1/86 1/86 TECHNOLOGY PLAN 14/8G 9/EE E/F.7 h/A PRA N/A 9/86 6/87 TBD f PSIL l
t
- OUTLINE COPPLETE COMPLETE COMPLETE N/A .
} - PSID StBilTTAL N/A 9/86 6/87 TPD
- LICENSABILITY STATEfEhT N/A f;/A C/F7 TFD 1
j 1
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4 O REVIEWSUPFCRT fiRF TECHNICAL ASSISTANCE MIT -
FUEL DESIGN AND PERF0PM CE ORNL - PERFORM INDEPEtIENT AfaLYSIS OF SPECIFIED ACCIDENTS AND TRAIGIEfRS. AID SPEB IN ITS REYlEWS AND ASSESEFENT.
Ef.L - SIMILAR TO ORNL, DEVELOPING A WATER / AIR INGRESS ACCIDENT ACCIDENT ANALYTICAL CAPABILITY.
I RES SUPPORT f
COMPLETION OF HTGR HANDBOOK:
O
I f0EEIGNTECHNOLOGY
~
l WORKING C0PtUNICAT10NS WITH ORNL
! THTR STARTUP AND HTR-500 DESIGN i INTERNATIONAL CONFERENCES i
1
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O O O J
i PLANT DESIGN OVERVIEW MODULAR HTGR PRESENTED TO THE ACRS JANUARY 30,1986 W.R. SHERIDAN - SENIOR PROJECT MANAGER
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l dSTONE & WEBSTER ENGINEERING CORPORATION l
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l DESIGN OVERVIEW OUTLINE I
KEY DESIGN SELECTIONS PLOT PLAN l NUCLEARISLAND ENERGY CONVERSION AREA L
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!. o o o OVERALL PLANT KEY DESIGN SELECTIONS FOUR (4) x 350 MWt MODULAR HTGRs TWO (2) x 300 MWe TURBINE GENERATORS SINGLE CONTROL ROOM, MULTI MODULE CONTROL SEPARATION OF NUCLEAR ISLAND AND ENERGY CONVERSION AREA I
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O O TURBINE CYCLE o ,
a g a , GROSS 1000*F GENERATION O 2415 PSIA HP LP LP LP LP -
,1005*F SIG ' '
1,5 ei TURSINE GENERATOR No.1 1 1 i
DEAERATOR p"' 25'HG q- ;p - -
CONDENSER 005*F Q ( ) FEEDWATER HEATERS 2 SIG t 2515 #2 2895 PSI A PSIA TURBINE GENERATOR No.2
,, , CONDENSATE PUMP i
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j 1000*F d L d b GROSS I 2415 PSIA GENERATION HP LP LP LP Lp -~
1005*F 300MWe j ' 2515 8#0 1 I 2.5*HG DEAERATOR p",' CONDENSER q ;;P FEEDWATER HEATERS
, 1005*F @ CONDENSATE PUMP
, 380*F 2515 S/G p4 ' 2895 PSIA
! I FW PUMP l O l STEAM i GENERATORS l
PLANT CYCLE PARAMETERS CORE POWER (4 x 350 Mwt) 1400 MW(t)
GROSS POWER TO TURBINE 1403 MW(t)
STEAM PRESSURE 2415 psia STEAM TEMPERATURE 1000 F TURBINE EXHAUST PRESSURE 2.5 INCHES Hg TOTAL GENERATION 600 MWe
- AUXILIARY POWER 42 MWe -
!; NET GENERATION 558 MWe .
l NET EFFICIENCY 39.9%
_ _ _ . , _ _ _ _ _ - - . . _ . _ , _ _ _ ___ _ _ _ . _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _