ML20155G142
| ML20155G142 | |
| Person / Time | |
|---|---|
| Issue date: | 10/05/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1695, NUDOCS 8810140103 | |
| Download: ML20155G142 (213) | |
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O UNrrED STATES l NUCLEAR REGULATORY COhDdISSION ADVISORY COM11ITTEE ON REACTOR SAFEGUARDS J
In the Matter of )
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SUBCOMMITTEE ON ADVANCED RFACTOR )
f DESIGNS )
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(Morning Session) )
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l DATE: October 5, 1988 LOCATIOM: v 0{Q9/-hM-Bethesda, !!aryla
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l PUBLIC tlOTICE BY THE 2 UllITED STATES 11UCLEAR REGUI.ATORY CORMISSIOtt'S 3 ADVISORY COKMITTEE Ott REACTOR SAFEGUARDS 4
5 6
7 The contents of this stenographic transcript of the 8 proceedings of the United States fluclear Regulatory ,
l
- 9 Commission's Advisory Committee on Reactor Safeguards (ACRS),
i 10 as reported herein, is an uncorrected record of the discussions I
11 recorded at the meeting held on the above date.
1 12 tio metaber of the ACRS Staf f and no participant at 13 this meeting accepts any responsibility for errors or
) 14 inaccuracies of statement or data contained in this transcript. [
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I Reporting Corporation O Heritage f (202) 628-4888 -
1 1 UNITED STATES NUCLEAR REGULATORY COMMISSION 2 ADVISORY COMMMITTEE ON REACTOR SAFEGUARDS 3 )
In the Matter of: )
j 4 )
)
] 5 SUBCOMMITTEE ON ADVANCED REACTOR )
- DESIGNS )
6 )
(Morning Sessiont 7 Wednesday, October 5, 1988 8
Room 116 9 7920 Norfolk Avenue Bethesda, Maryland i 10 The above-entitled matter came on for hearing, 11 pursuant to notice, at approximately 8:30 a.m.
- 12 i
BEFORE: MR. DAVID A. WARD
. 13 Research Manager on Special Assignment
- E.I. DuPont de Nemours & Company
() 14 Savannah River Laboratory Aiken, South Carolina
- 15
, l' i_liEMBERS PRESENT:
l 16 DR. WILLIAM KERR 17 Professer of Nuclear Engineering and Director of the Office of Energy Research 18 University of Michigan Ann Arbor, Michigan 19 DR. FORREST J. REMICK 20 Associate Vice Fresident for Research and Professor of Nuclear Engineering 21 The Pennsylvania State University University Park, Pennsylvania 22 HR. CHARLES J. WYLIE 23 Retired Chief Engineer Electrical Division 24 Duke Power Company Charlotte North Carolina 25
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1 DR. CHESTER P. SIESS .
l Professor Emeritus of Civil Engineering 2 University of Illinois O' ' Urbana, Illinois 3 i HR. JAMES CARROLL 4 Retired Manager, Nuclear Operations Support ;
Pacific Gas & Electric Company I 5 San Francisco, California i 6 HR. CARLYLE HICHELSON i Retired Principal Nuclear Engineer 7 Tennessee Valley Authority .
Knoxville, Tennessee 8 and Retired Director, Office for Analysis ;
and Evaluation of Operational Data l 9 U.S. Nuclear Regulatory Commission l Washington, D.C. !
10 U
CollS_U_LTANTS : ,
11 ,
R. Avery 12 4CRS COGJIZAtJ_T I STAFF HEt!B_ER:
13 Medhat El-Zeftawy i(RC_ STAFF PlESENTERS: ,
15 King '
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16 R. Landry 17 !
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{) 2 CHAIRMAN WARD: Good morning. The meeting will now 3 come to order. This is a meeting of the Advisory Committee on 4 ' Reactor Safeguards Subcommittee en Advanced Reactor resigns.
5 I am David Ward, the Subcommittee Chairman.
6 The other ACRS members in attendance are Hr. Kerr, 7 Mr. Michelson, Mr. Carroll, Mr. Remick, Mr. Siess, and Mr.
8 Wylie. We also have our ACRS consultent. Mr. Avery.
9 The purpose of the meeting is to review the draft 10 SER for the PRISH design. That's the Power Reactor Inherently 11 Safe Module. Medhat El-Zeftawy on my right is the cognizant 12 ACRS staff member for this, for the meeting.
13 Rules for participation in the meeting were
() 14 announced as part of. the notice of the meeting published in 15 the Federal Register on September 22nd. A transcript is being 16 kept, and will be made available as stated in that notice. I 17 request that each speaker speak loud enough and with clarity 18 so that he or she can be readily heard. That should mean 19 coming to one of the microphones around the room.
20 We have a--let's see. All of the Subcommittee 21 members and Mr. Avery should have received in advance a couple 22 of weeks ago a meeting summary and a copy of the draft SER.
23 Did everybody get that?
24 HR. WYLIE: Yes.
() 25 CHAIRHAN WARD: We have a couple modest changes in HERITAGE REPORTING CORPORATION -- (202)628-4888
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l 1 the agenda. We are going to, staff has asked that we switch 2 the order of items II and III so that Mr. Jerry Griffith of 3 DOE will be our first presenter, and Tom King of NRC Research ,
4 will follow.
5 Also this afternoon I understand that item V, the 6 ANL fuel behavior study, that there will not be a presentation 7 from ANL people on that, but there are people here to answer 8 questions that we might have in any discussion of that area. j i
9 Do a*4y of the members have anything they would like ;
10 to say before we begin with, we go to Mr. Griffith? Okay, f
11 Bub, did you have anything? No? Okay. Then I ask '
12 Mr. Jerry Griffith of DOE to begin his presentation.
I 13 HR. GRIFFITH: Thank you. I'm pleased to have the
() 14 opportunity to discuss DOE's LHR program with you this 13 morning. I believe that these meetings are very important 16 for reasons that I will indicate in the discussion, and I +
17 would like to start by indicating a little bit of background 3 18 of where we have come from, where we are, and where we think l
19 we are going, so I will cover a little bit of that background, f 20 the strategy that the program is based on, the description of 21 the program, and finally a little bit about the budget. i i
22 (Slide) 23 HR. GRIFFITH: First, I would like to indicate that -
24 energy usage has held steady. After the shocks in '74, the ;
25 total use of electricity except for a period of recession in i
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1 5 l 1 the early '80s, has continued to climb, and follows the Gross 2 National Product which has been a traditional trend, and we 3 expect electricity usage to climb further in the future.
4 Further historically, as is indicated in that chart, 5 the percentage of use of electricity has continued to increase 4 6 from 14 percent in the '50s to presently in the 30s, and by 7 the year 2,000, to 40 percent of total energy production and 8 use, i t
9 (Slide) 10 HR. GRIFFITH: In recent years, the addition to i
11 capacity since '80, is indicated here--coal, 30 percent, ;
4 12 nuclear, 58 percent, and they are the, by far the two most !
I 13 used fuels for electricity production. !
() 14 (Slide) 15 HR. GRIFFITH: During this same time, the public P
j 16 acceptance toward nuclear power has steadily decreased from, ,
17 as shown in this chart. THI and Chernobyl, while probably 18 affecting public attitudes, have not been clearly the 1 ;
19 contributing, major contributi:.g cause for this discomf ort on i !
20 the part of the public with nuclear power.
21 (Slide) 22 HR. GRIFFITH: As a result of this public
! L i
23 discomfort, and the operational and regulatory difficulties t
1 24 with these plants, the cumulative benefits that was accruing 4 i 25 from nuclear power beginning in the early '70s, in the early
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6 1 '80s began to decline and actually decreased such that now as
() 2 nuclear power is used, comes on line, the benefits to the 3 public are negative.
4 (Slide) 5 HR. GRIFFITH: So what we have is a nuclear 6 dichotomy in the United States. The performance of 109 plants 7 licensed to operate, has been generally good, and capacity of 8 97 plus megawatt electricity, and there has been an ex7mplary 9 safety record in spite of the fact that there are a few 10 problems that should be and can be fixed. There have been no 11 plant orders in the last ten years, so that's where se are.
12 (Slide) 13 HR. GRIFFITH: The Energy Information Agency at the 14 Department of Energy predicts nuclear usage in the future as 15 indicated here, that beginning in the year 2000, three cases, 16 no nuclear orders, we would decline dropping around 2005, a 17 lower reference case could, slight increase in an upper 18 reference case in which there could be a considerable 19 contribution of nuclear power. We expect that the LHR can 20 provide part of that power that is indicated, and that if it 21 does that, there would be benefits that I will describe later 22 in the presentation.
23 A little bit about the program strategy then--most 24 of you, I'm sure all of you in this room are aware of the plan
() 25 that was in effect in the late '60s and early '70s in the HERITAGE REPORTING CORPORATION -- (202)628-4888
I 7 i
- l 1 Atomic Energy Commission to build a breeder. There was a !
[} 2 demonstration program for LWRs. There was a growth in LWR t a 3 usage, and in 1974, these were the projections that were made !
, [
! 4 by the then Atomic Energy Commission, that LWR usage would i I climb as indicated here, that this would put pressure on 5
4 !
6 uranium prices as the high grade ores were mined, and [
s j 7 eventually, the breeder and reprocessing would be economic, f I
8 and that the breeder after demonstrations would be coming into -
f
- 9 the, our inventory in the mid-90s, at which time the i ;
i 10 converters and the breeders would form the backbone of our l k
1 11 future energy supply.
l 1 ;
1 12 (Slide) I E
- 13 HR. GRIFFITH: That did not happen. Realities of {
() 14 the '80s are Clinch River has been cancelled. There is no
- 15 commercial breeder demonstration program. The current DOE i
16 forecast indicates a flat usage as indicated in the previous
[
17 vuoraph for nuclear power through the year 2000, and this [
t i 18 occurred for a number of reasons. Some of them are listed on (
i 19 this chart, but the two primary reasons are that the growth 20 rate of electricity consumption was not as high as was i
21 indicated or expected in the '70s when growth rates were about i i
! 22 7 percent. In the later years they were more like 2, maybe 3, i I .
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23 although they have climbed back up and are in the range of 5 !
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24 percent now. l l< '
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() 25 The second major reason was that more uranium l
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8 i a 1 resources were found than was originally anticipated. As a l i
2 consequence, the need for the breeder has receded well into 3 (
i 3 the future. Dates like 2030 and beyond are discussed. .
! 4 (Slide) r 5 MR. GRIFFITH: However, we--at the time the breeder 6 program was in this status, over $7 million had been invested !
l j 7 in liquid sodium technology and as you know in this room, the l 8 first reactor that was operated in Idaho was the liquid metal, f
-l 9 and this had been an extensive operating experience. The 10 thought that obviously occurs is what other use this 11 technology can be put to besides breeding? And we were !
i 12 looking at some problems faced by the LWR industry, and the l
] i 13 question comes up can the LMR technology be used to make power f t
() 14 plants that are economically competitive independent of their ;
15 breeding capability, and would they provide a technology that l 6
1 16 might make easier solutions to some of the problems facing the l i
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- 17 nuclear industry? I f
j 18 I won't go through these. They have been discussed i i l 19 in detail, but there is an interrelated problem of regulatory, f
] e 20 public attitudes, the complexity of the plants, that has l 1 i
- 21 grown, their operating complexity. There is uncertain load !
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l 22 growth, and financing problems particularly that e.re faced by j l 23 the nuclear industry because it is a high capital cost f l 24 industry, and if a product is to be better, it has to address !
25 these, these concerns and these challenges in a systematic
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1 manner.
2 We took a look in this program as well as in our 3 other advanced reactor program, the gas cooled reactor, and we ;
i 4 looked at these challenges, and we asked ourselves how can we l 5 better meet these, and we arrived at four, four major items in i
6 the LHR that needed to be examined and addressed and offered a l .
I 7 good possibility of improvement. They are listed '
8 here -passive safety modularity, standardization, and improved l
9 waste management. I will talk to those a little later.
l 10 Standardization I won't say much about. That's with LWRs all l 11 over the world it become clear that we must do better in this 12 area. I will talk a little bit more about the modularity 13 later, but first I would like to say a few words about passive 14 safety.
t 15 Why are we interested in passive safety? First, we I l 10 think it inproves safety margins as well as increase, enhance i i j 17 the transparency of the safety margins.
, t 1 :
i 18 DR. KERR: What is a safety margin Hr. Griffith? l I
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- 19 HR. GRIFFITH
- Pardon? f 20 DR. KERR What is a st,lety margin?
l 21 HR. GRIFFITH: Safety margin is the, the distance i i
f 22 between the operating limits and the point at which there
? I l 23 would be a problem that could lead to public consequences, i
' l 24 DR. KERR: If I say it improved safety, I would be O s v1=e the tai #at 1 taere o etsi=2 aittere=t aetwee= !
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1 2 HR. GRIFFITH: Well, I think there is much, I think !
1 3 that improving the safety margins, there is a lot of ways to :
j 4 improve safety. I think improving the safety margins is one, l
! i i 5 and that's what this does. I mean you can work on quality l 6 control. You can work on operator performance. There is a ;
! 7 lot of things you can do, and in this case, one of the things I 4
8 is to improve the margins between tue damage points and the I <
l 9 operating points. !
l j 10 DR. KERR: You can improve safety margin without (
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- 11 improving safety?
l 12 HR. GRIFFITH: I think you could. I think you would f
i i j 13 have to unimprove it some place else, p
() 14 We think it leads to simpler plants and simpler
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15 operations because of this, and that the risk of core e 16 accidents can become insignificant, and this could aid in !
17 plant capital investment protection and aid in public !
18 perception of safety, but the buttom is we think all of this !
l 19 above leads to plant cost reduction when looked at in an ;
i 20 integrated way, t 21 CHAIRMAN WARD: Gary, the core melt risk, core melt 22 being insignificant, is there some threshold you assigned? I 23 mean what do you mean by insignificant? And is there some 24 quantitative expression of that? And is there some way that f
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(} 25 you believe you can demonstrate that it is insignificant by f
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t 11 1 whatever measure you use?
{} 2 HR. GRIFFITH: Let me address that in two ways. I 3 want to move from here to the, to the HTGR program because I -
- 4 think it is even more transparent. ;
5 In that case we believe there is no way to melt the l 6 core down because the design was conducted with a requirement I
J 7 that that be so, and the plant in its--will dissipate decay j j heat by conduction, even if all other methods of heat transfer 8
! 9 have been lost in the plant, and the reactivity coefficients l 1 l 10 have been done such that we do not see any identifiable way !
i r i 11 that the core could melt down. !
i l 12 Now the LHR is almost there with the containment, I j i j 13 that is secondary containment that is put around the coolant, f
() 14 and again using passive safety reactivity design techniques, '
15 et cetera, that plant also is extremely safe, but cne I think l 5 f
} 16 has to look at some probabalistic assessment of losing the i r
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- 17 coolant, and we believe these are extremely low, certainly [
! 18 well below the NRC guidelines for the LWR and the advanced {
! i l 19 reactor safety goals. !
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[ 20 We have analyzed them, and that-- f 1 1
! 21 CHAIRHAN WARD: When you use core melt, of course f 22 what we are really interested in is the release of fission r
I 23 products, and for the HTGR, for example, it might be closer to i r
I 24 the truth to say there is absolutely no way the core can melt !
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() 25 than it is to say there is absolutely no way the fuel fission 1 i
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12 1 products--
2 HR. GRIFFITH: I misspoke. The way it is designed, 3 there is absolutely no way to release the fission products.
4 The temperature involved in the conduction of the heat away, 5 you would lose some of the plant investment, but it would not 6 release fission products.
a 7 (Slide) t
- 8 HR. GRIFFITH
- Okay. In the LHR, the way that was
- 9 chosen within the design to obtain passive safety enhancement i 10 is with the metal fuel and IFR technology, modularity of l
j 11 course, advanced instrumentation and controls to handle the 12 modules, r.andardization. and again, the improved waste l
1 j 13 management that comes from the IFR fuel, metal fuel cycle and
() 14 a potential for actonite burning. I will speak to that more 15 later.
]
16 DR. KEHICK: How does tne IFR technology contribute 6
l 17 to passive safety?
l j 18 HR. GRIFFITH: Play a straight man for me. The j 19 metal fuel cycle potential benefits are outlined here. Very 1
l 20 quickly, I would like to say that in the days in the '70s when
- )
j 21 oxide fuel was selected as the reference for the breeder
- 22 program. I believe that decision was clearly the right one.
l
'3 At the time, there was experience with oxide fuel from the I
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'4 water reactors, and there was an industrial capability, and I
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{} 25 the metal fuel had a swelling problem and a few problems that
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l 1 had not been solved, so at that time, it was decided to go 2 with tho oxide fuel, and it has been developed to a very high j 3 state of reliability, and we have completed that development !
, i l 4 program primarily with the Japanese cost sharing 50/50 on the ;
- 5 remaining programa. !
6 However, if one were to look at metal fuel, which I
) !
i 7 was some of the earlier fuel used, the comparison to oxide, [
I l j 8 there are some clear benefits. The oxide fuel is not !
l i j 9 compatible with sodium leach as if there is a breach in the !
I f 10 clad. The metal fuel is sodium bonded, completely compatible i
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f 12 high center line temperatures, and in many transients that ;
- c l 13 feeds back adver'iely to the transient, doesn't make positive i
14 coefficients, but it does adversely affect the acility to I
j 15 control. !
l 16 And finally, the oxide fuel is not as dense as metal !
j 17 fuel and therefore, does not breed as well, and all of these ,
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! 18 things can be traded in the design for econom!c benefits and l
} I 19 improved safety margins. [
l
- i l 20 DR. REMICK: But that answered the question why i
! I j 21 metal fuel. That wasn't the question. That I understand.
f 4 ;
22 The question is how IFR technology contributes to passive l 23 safety? I i
24 HR. GRIFFITH: I think that one of the principal f t,
t 25 areas that is being worked on at Argonne National Laboratory f
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14 4 1 is the inherent control techniques that are being looked at
[} 2 using EBR 2 as a development tool. They are examining the 3 possibility of operating the pump completely on the pumps, 4 with the response of the reactor slave to the, to that 5 requirement.
6 The other, I think that the IFR technology is a kind 1
7 of a, it is an enveloping word that describes all of the more i
8 fundamental and basic research that is being done by the 9 laboratory that helps feed into the design activities of the
. 10 ven.ers, and to the extent they are looking at any advanced l i
11 developments that are of this type, that's the word that, ;
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! 12 that's what the IFR technology is meant to describe, but the !
1 13 principal part of that IFR technology is the metal fuel cycle.
)
( 14 DR. REMICK: Okay. It has nothing to do with the
! 15 possibility of reprocessing or the integral cycle or anything i 16 like that?
17 MR. GRIFFITH: It does have.
- 18 DR. REMICK
- How does that contribute to passive
! 19 safety? How has IFR technology contributed to passive safety? i
( 20 There is something I am missing. I can understand the fuel, !
21 but that's one of your-- ,
L 22 MR. GRIFFITH: That's half or more of the IFR I l
23 technology.
24 DR. PEHICK: All right.
() 25 CHAIRMAN WARD: But I guess the point is your chart f
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i 1 wouldn't have lost any accuracy if you had under passive--the 2 previous one--if under passive safety you had just said metal 3 fuel. And the IFR technology perhaps makes it practical to 4 have metal fuel, but it doesn't make.a separate contribution i
5 to passive safety of the reactor that I can see. (
6 HR. GRIFFITH: Well, let me take one other item. -
7 For example, in the pump, electromagnetic pump, we are looking j 8 at Argonne to develop a wire insulator that has high 9 temperatures such *. hat the, that the pump can be cooled by the 10 flow of the pump sodium. Such a development would elimin**a r'
11 the need for a cooling system. That cooling system can, 1. At ;
12 were there, can cause maintenance problems. It could cause r
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13 potential mixing of two different kinds of fluids. There are things that can happen in the plant if that system is there,
(]) la 15 and we feel that the simpler the plant is, the safet the plant i
10 is; and any of these things that we can develop will
, 17 contribute to the safety of the plant.
18 And I don't want to quibble with words, but f l P j
19 basically we have two kinds of efforts at Argonne. One is j 2 !
j 20 metal fuel cycle, which is very well described.
I i 21 The other is some of these other kinds of activities 22 that can lead to the simplification of the plant design and i
j 23 the plant operation. [
t t
24 DR. KERR: I must say I find arguments that t
25 simplicity makes things safe not very convincing. Simplicity J
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16 1 that makes things safe, makes things safe, but there is
'T 2 sometimes complicated systems that are required to make things (G
3 safe, and just to tell me that you are simplifying something 4 and that makes it safer I find extremely unconvincing. If you 5 can demonstrate why making it simpler makes it safe, I am 6 willing to 11 co that, but the argument just doesn't hold 7 water.
8 MR. GRIFFITH: We think simplicity is one--I may 9 hhve misspoke again. Let me say we think simplicity ir one 10 technique that one can use to enhance the safety of the 11 plants. I gave one example. I think when you have got two 12 fluids floating around in a plant, it gets difficult to design i
13 and maintain separation. They mix all the time. It happened 14 in the past.
15 DR. KERR: If you could demonstrate that simplicity 16 makes things safer, then you know, do that, but just to say 17 that because you are making something simple, it is therefore 18 safer, is not a valid argument.
19 HR. GRIFFITH: I hope I haven't said that. If I 20 did, I take it back.
21 In the fuel area, we can make smaller cores. We 22 found in our designs that with the high degree of internal 23 conversion that we can build a reactivity in the core that 24 needs to be controlled far less than a dollar, and that leads
() 25 to less concerns about reactivity control accidents, and of HERITAGE REPORTING CORPORATION -- (202)628-4888
17 1 course, the improved breeding potential is there, but one of 2 the things that we found that has a large advantage for us, O~
3 maybe less to safety, was the improved fuel cycle economics.
4 It turns out that these, the metal fuel processing 5 are kind of hot box, glove box kind of operation that can be 6 conducted in batches that are economical, in very small 7 volumes, and compared to the aqueous solutions, offer large 8 benefits, maybe factors of four or five, in the volumes and 9 costs of the reprocessing, and I give an example here that 10 when we were looking at the oxide reprocessing we were looking 11 to build at Hanford, we were talking about investments of $300 12 million or so, and these kinds--we are now looking at 13 modifications to HFEF South at Argonne that will be more in
() 14 the tens of millions, 15, 25 million dollars, that will allow 15 HFEF to process all the metal fuel that would be required for 16 EBR 2 for FFTP, and the first commercial plant for, of about a 17 thousand megawatts, 18 This is really a good comparison because the HFEF 19 South wasn't there, but it does give indication of kinds of 20 cost at least to do this, and they are very expensive, and so 21 we are quite intrigued with this opportunity to close the fuel 22 cycle without a large investment with the metal fuel course 23 that we are taking, and I, we have always in the past been 24 faced with the first breeder that the industry might have to 25 build either being with a fuel cycle that is not closed, or HERITAGE REPORTING CORPORATION -- (202)628-4808
18 1 with this large investment, and this is a tremendous advantage
() 2 for commercializing this technology.
3 CHAIRMAN WARD: There certainly seem to be some 4 advantages here, but if a utility is to build one or a lot of 5 these things or a lot of utilities build one or more, the 6 utility is going to have the job of running this fuel process, 7 the.IFR process as well as running the reactor, which doesn't 8 sound like a simplification of the utility's task.
9 What, do you think that might be a problem, or are 10 there going to have to be different kinds of utilities to 11 successfully manage this whole thing?
r 12 MR. GRIFFITH: Well, I think it is not a problem.
13 We--IFR does not include the requirement for an integral i 14 reprocessing capability on site. Argonne, and we both agree 15 on this, and when they talk about IFR and we talk about IFR.
4 16 that does not imply that it has to be on site, and we do not 17 see that there is any--if you want to make it centralized, it
- 18 turns out that you will build it in batches, modularized, even 19 if you did it centralized, so you lose nothing. You could l 20 centralize.
21 Second of all, there are companies, and Bob Berglund 22 is here from GE. They have stated to us in the past that they i
23 would be willing in the future to look at such organization of 24 this that they would do it contiguous to the utility operation l
() 25 such that the utility did not have to be concerned with this, HERITAGE REPORTING CORPORATION -- (202)628-4888 l
19 1 so I don't think it would be a problem.
2 CHAIRMAN WARD: I see, so you wouldn't necessarily 3 be shipping spent fuel to processing plants? You might be, 4 but you wouldn't necessarily, but you might have a specialist 5 in processing build . plant next to each one of these.
6 MR. GRIFFITH: Yes. Now I think that's the real 7 problem. I believe that if one looks at the, at the real 8 risks of releasing fission products to the environment, that 9 one would probably right now be much more concerned about that 10 part of this operation than the reactor, and you do put that 11 additional risk at the site at the time you do it instead of 12 some place elso, but then you don't have to ship, and I think 13 those questions need to be looked at very carefully.
() 14 CHAIRMAN WARD: Thank you.
15 MR. GRIFFITH: Okay. In the designs that we have 16 looked at then using these techniques, we have selected PRISM 17 as our reference design to study further, and the only point I 18 want to make then on this vugraph is we have attempted to make 19 the reactor such that it is not influenced by anything that 20 happens in the balance of plant, the steam generator rating 21 part of the plant, and therefore, that the possibility to make 22 the balance of plant to high grade industrial standards as 23 opposed to safety grade exists, and if such a thing could be 24 done, there could be considerable savings on the plant.
rg 2S DR. KERR: What is high grade industrial standard?
U HERITAGE REPOR1ING CORPORATION -- (202)628-4888
= J & -* . 4 .- 4 1 a k
20 ,
1 HR. GRIFFITH: You can take the same steam plant, 2 the coal plants, put that in as a base. You can take it--it 3 can't affect the safety of the reactor.
4 DR. KERR: Is that what you mean by high grade 5 industrial standard?
6 MR. GRIFFITH: Well, that or augmented from there.
7 DR. KERR: I am just trying to find out what it is 8 you are referring to when you use the term high grade 9 industrial standard.
10 HR. GRIFFITH: It would be a plant like a coal 11 plant, augmented as required to, to meet any special problems l 12 that one finds in a design.
13 DR. KERR: Where would one find a high grade r
() 14 industrial standard? Does it now exist, or would it have to f
15 be developed? -
l 16 HR. GRIFFITH: I think it presently exiscs. i 17 DR. KERR: Thank you. I 18 CHAIRMAN WARD: Why does it, why does it have to be 19 high grade?
I 20 HR. GRIFFITH: Well, I think you want the plant to i
21 be, I think that compared to a coal plant, you have a higher
- 22 capital cost plant. You have a lower cost fuel hopefully, and l
23 you want it on line, so I think that compared to a coal plant j 24 you would take the best they had to offer and use it in this l
{} 25 plant again as a first principle.
HERITAGE REPORTING CORPORATION -- (202)628-4888
21 1 CHAIRMAN WARD: So but as far as let's say the 2 regulator is concerned, or as far as someone worried about the 7
U 3 reactor safety is concerned, whether it is high grade or not, 4 it is, is unimportant. I know this is sort of an extreme 5 position, but I mean I am trying to see whether you really C believe it is not part of the safety.
7 MR. GRIFFITH: One must get down to the specifics 8 and they must examine them, but for example, failure of 9 feedwater is not a problem for the plant or the reactor.
10 CHAIRMAN KERR: The light water reactor people even 11 thought that for a while. Since it didn't make feedwater 12 safety grade, not even aux feedwater safety grade. They did 13 that for a long time, as a matter of fact.
() 14 MR. GRIFFITH: We have learned a lot since that, 15 these first reactors, and we have learned that the balance of 16 plant was influancing the LWR reactor, and the design in this '
17 after that 25 years of learning or so, we have designed the 18 plant such that it is not--and one has to look and see did wo 19 or didn't we.
20 DR. KERR: Let me repeat that the light water 21 reactors were designed so that feedwater system doesn't
. 22 influence the plant. In good faith, everybody presumably 23 thought $his.
24 What leads you to believe that yot have been any 25 more successful particularly with an experien e that you have HERITAGE REPORTING CORPORATION -- (202)628-4888
22 1 with this plant?
/"N 2 HR. GRIFFITH: I don't believe that in the latter
(-)
3 days of the LWR it wac accepted that any.
4 DR. KERR: I am not talking about the latter days.
5 I am talking about the stage of design in w:ich you now are 6 with this plant.
7 HR. GRIFFITH: Well, in the early days of 8 regulatory, the decisions were very deterministic, and there 9 was no analysis as to how much the balance of plant could 10 affect these things.
11 In many cases, and as we became more sophisticated 12 in these analyses, it was found out that they were the first 13 lovel of defense in safety.
() 14 DR. KERR: As I read the SER provided to us by the 15 NRC staff. I find that the staff at least has concluded that 16 they have to be deterministic in most of the decisions that 17 they make at the present time because tha uncertainty in 18 probabalistic estimate is so high.
19 HR. GRIFFITH: That's right, becau3e the 20 probabilities are so low.
21 DR. KERR: That isn't the whole story. That 22 probabilities are low, but the uncertainty is very high, which 23 means that it may be low and it may be considerably higher 24 than those calculations indicate.
(} 25 HR. GRIFFITH: There is nothing wrong with doing HERITAGE REPORTING CORPORATION -- (202)628-4388
23 1 deterministic, but there also is nothing wrong with looking at 2 the balance of plant, and analyzing if you could cut that 3 entire thing off, and if there was any influence on the 4 reactor.
5 DR. KERR: All I am trying to say is I don't see how 6 you can be so confident about these things at this stage of 7 the development of the plant, particularly when one looks at 8 past experience.
9 MR. GRIFFITH: Well, I'm not here to try to license 10 this plant right now, and I'm not saying our confidence is 100 11 percent until we have interacted with the regulatory 12 authorities and others to take a real hard look at this, and 13 that's exactly why we are here.
() 14 I'm here to say what our approach has been to try to 15 satisfy the constraints of safety regulation, public 16 attitudes, and economics. This is cur approach. We are 17 trying to use these techniques to the maximum extent possible 18 and to the extent that we could possibly eliminate the balance 19 of plant from being a safety concern on the reactor, that's 20 what we are trying to do.
21 I'm not standing here and telling you we have done 22 it. We are here. We have got people. We will have 23 interactions in the coming years, and hopefully we can 24 convince you.
25 CHAIRMAN WARD: But this is your goal?
)
HERITAGE REPORTING CORPORATION -- (202)628-4888
24 1 HR. GRIFFITH: This is our goal, and I'll tell you--
2 CHAIRMAN WARD: So if 50 years from now there are
{T s-)
3 two PRISH plants operating side by side, and one is having a 4 lot of balance of plant failures and outages and the other 5 isn't, you would expect the NRC to be giving them similar SALP 6 ratings possibly, or not necessarily dissimilar SALP re. tings?
7 HR. GRIFFITM: In extreme, yes, but the rate payers 8 won't.
9 CHAIRMAN WARD: I agree with that, but the NRC and 10 the rate payers have, should have two different perspectives.
11 Could I ask you one other question? You said that 12 you selected PRISH as a reference design for the, for your 13 LHR program. We have had some confusion about that here. Is
() 14 that, that is the final decision and the SAFR effort has been l
15 dropped?
16 HR. GRIFFITH: Yes. That's the final decJsion. Wo 17 would like the NRC to wrap up the investment that they have 18 put into the review of SAFR, and give us feedback on that 19 design approach because there may be some of that that will be 20 applicable as the PRISH design is finalized, and it would, it 21 would help to have those views, especially since the work has 22 been done and involves kind of like just putting out a report.
23 CHAIRMAN WARD: Yes.
24 DR. SIESS: Was the decision to go with PRISH based l
() 25 on safety considerations or other safety considerations?
l l
l l HERITAGE REPORTING CORPORATION -- (202)628-4888
25 1 MR. GRIFFITH: It was based on a large number of 2 things. I do not believe that safety was an overriding 3 consideration because I believe we believed that both of these 4 plants were extremely safe or could be made extremely safe.
5 DR. SIESS: Did you f eel that the straf f shared that 6 opinion?
7 MR. GRIFFITH: I believe so.
8 DR. SIESS: Equally safe?
9 MR. GRIFFITH: Yes. They are here. I would ask 10 them. It is my impression--
11 DR. SIESS: I was just wondering. The point had 12 been made in something I read that you thought there had been J 13 enough interaction with the staff to take into consideration 14 their concerns of making your choice between PRISM and SAFR.
(])
15 Has that interaction given you any hints they 16 thought one was more safe than che other or they would have 17 less trouble with one than the other from a safety point of 18 view, or whether they were running neck and neck?
19 HR. GRIFFITH: That was my understanding.
20 CHAIRMAN WARD: Well, one, one might have concluded, f 21 you know, in your making the choice, that either the choice 22 wasn't made on regulatory safety issues, or that you had 23 enough informal work from the NRC staff as to their opinion on 24 regulatory safety issues that you felt comfortable or 25 confident in making the decision.
1 l
i HERITAGE REPORTING CORPORATION -- (202)628-4888 I
26 1 Maybe--I think we will hear from the NRC staff later
- 2 with their opinion on this. It did seem a little let's say
(~j 3 odd that the decision was made before the NRC staff, you know, 4 supplied you with their formal paper. You know, maybe your 5 informal communications were so good that it wasn't necessary.
6 MR. GRIFFITH: We would have preferred otherwise, 7 but we were faced with a discontinuity of losing the 8 contractors over contract arrangements and budget, and the 9 deputy assistant secretary was forced to make a decision and 10 he had to do it on the best information and data he had 11 available, and there were other reasons for the selection of 12 PRISH with the understanding that we did not see any show 13 stoppers at the NRC for either of the designs, and by show
(') 14 stoppers. I mean something that either wasn't, that was agreed 35 to or seemed to be agreed to or couldn't be fixed in some way, 16 in places where NRC was nervous.
17 The balance of plant contains 80 percent of the 18 plant cost in existing LWRs, and the economy, if we are going 19 to look at modularization, which contributes a lot to l
20 financial ability to match load growth and finances, and 21 improve some of the safety margins, then we must gain some 22 place else, and this was one of the places we thought we could 23 do it. We have analyzed our designs compared to coal and 24 current LWRs and they fall in the range that is competitive.
25 And so we are convinced that this approach has some HERITAGE REPORTING CORPORATION -- (202)628-4888
27 1 merit and can work and we are aware that, that paper reactors 2 are always more simple, at least less costly and all that, 3 than real reactors, but we have tried hard with peer reviews 4 and other techniques at our disposal to make sure that we 5 aren't fooling ourselves. For example, we can look at the 6 commodities that are in these plants, and there is less 7 commodities and clearly less cost in some of these areas and 8 some of these are listed here.
l 9 Finally, there is another aspect, and that is with I 10 this kind of approach that we have used, we find the 1
11 distributions of costs compared to current technclogy, that j 12 there is a different distribution. Where lasss of the cost is l
13 an indirect and more is in equipment and materials, the
(/ 14 customer utility has more control over equipment materials 15 that he can get on fixed bid than he does the indirects which 16 are related to site management, A&E costs and other things.
17 CHAIRMAN WARD: Is there some, can you give us, 18 well, an example or some overriding reason why the indirects 19 are so much different? I mean what is the nature of those 20 costs and the nature of the designs?
21 MR. GRIFFITH: Well, I think the biggest thing 22 probably is the greater extent of factory fabrication, and I !
23 think also, though, that the standardization which the 24 comparison there to an existing plant is probably not as 25 great, but I would believe standardization, factory
({}
HERITAGE REPORTING CORPORATION -- (202)628-4888
28 1 fabrication are the largest contributors to that, and they
(}
r 2 aren't that hard to obtain.
3 (Slide) 4 HR. GRIFFITH: Now I want to give one example of why 5 it is important that we have early interactions here in our 6 design approach.
7 One of the things that we talked about was if the 8 balance of plant could be not safety grade, there would be 9 cost improvements there, and we wanted to look at where we 10 were on this.
11 Bechtel did a study in which they looked at PLBR 12 pool. They looked at a Superphenix replica and LSPB loop 13 plant, and LSPB pool, and finally the modular reactors that we 14 are looking at here and what we are looking at here in a ratio 15 of LHR to LWR plant costs on the vertical axis. The 16 horizontal action is an LMR to LWR nuclear steam supply rttio, ,
17 and finally the diagonal line is the balance of plant ratio, 18 LHR to LWR, so one can look in here at the designs as they 19 analyzed them, as Bechtel looks at the water, and/or the 20 former balance of plants, modular reactors down to about .65 21 so there have been gains and costs of the balance of plant as 22 we expected, from the simplification of interaction with the 23 primary or reactor system, but we find something else on here, 24 that while those costs have gone down, the cost of nuclear
() 25 steam supply has stayed about constant at about 3, aid the HERITAGE REPORTING CORPORATION -- (202)628-4888
I 29 1 question naturally comes up why is that so? And it turns out
,_ 2 that the low pressure system that sodium offers cannot be
\/
3 taken advantage of because seismic constraints hit you before 4 you can make this smaller and less beefy equipment, so the way 5 we looked at what could be done to do that, or to improve that 6 situation, and again Bechtel did this study with Japanese and 7 DOE funding, about a million dollars, they looked at a 8 different way of supporting a reactor vessel of PLBR type, Gnd 9 found that with a new kind of bottom support approach to 10 seismic for these pool reactors, they could make savings of as 11 indicated here, 40 to 70 percent in some of the related 12 components, with a 58 percent savings of about S30 millions 13 for those components that were involved.
(} 14 Now this is not the seismic approach that we would 15 eventually colect for SAFR, and SAFR is using an isolation 16 approach that is more innovative, and we don't have a way to 17 direct compare with something in the past like LSPB, but we do ,
18 need then to examine these things as they go, and the 19 approaches that are new and unique need to be discussed very 20 early with the regulatory authorities so that the regulations i 21 can either be accommodated or they can be improved if that's 22 possible to, to meet those design requirements.
23 (Slide) 24 HR. WYLlE: All of your cost figures are based on 25 what construction cycle, time?
O HERITAGE REPORTING CORPORATION -- (202)628-4888
30 1 HR. GRIFFITH: I believe the construction time of 2 the modular plant was four years.
3 HR. WYLIE: And the figures you used for the L 4 conventional light water reactors is what?
5 HR. GRIFFITH: I believe they are eight to ten.
6 HR. WYLIE: That's a major difference in your cost
! 7 figures?
8 HR. GRIFFITH: That is one of the advantages of 9 modular factory fab, build them. One thing we have tried to 10 do a lot more of here is get some effective international I 11 collaboration and we have some fairly extensive programs with 12 Japan especially. The design approaches in these countries 13 are not the same right now, but there is still a lot of things Q 14 that we can do, and I, in this case, I will pick one of the i 16 more important items off here and discuss it a low minutes, !
t 16 and that is the completion of the oxide fuel development, t
l 17 In this case, we used to spend something like $20 18 million a year at Oak Ridge working on reprocessing technology 19 development. We have reduced those costs to the departments ;
I t
- 20 to 5 million, and Japan pays 5 million, so that the program
( 21 for five years of $10 million will complete that, and the t
{ 22 Japanese acquire access to our paist development technology, we
! i 23 acquire access to the use of that technology in their l t
24 processing plant at Orai, and therefore both of us at much ,
25 reduced cost to both of us, will be able to have this I l O :
l HERITAGE REPORTING CORPORATION -- (202)628-4888 u_ _ . _ .__ _ _ _ _ _ _ . _ _ . _ .
31 1 technology in the future if it is needed.
{} 2 Now we are interacting also with Japan to get them 3 involved in the metal fuel cy:le development, and they are 4 very interested, and our venders are interacting to try to 5 bring a commonality of component design and other arrangements 6 that are more design related, and I think that there will be 7 much advantage in the future of continuing to build these 8 kinds of international arrangements.
9 And again with all this lead time increase on some 10 of these things, the importance of early interactions and 11 knowing where we are going is very important, so finally, the 12 advanced program is, has four major components--one, 13 completing the off-site cycle as expeditiously and
(/ 14 economically as we can, developing the metal fuel cycle to see 15 if the potential that we think is there is, and finally, the 16 additional technology development to simplify our plant 17 designs, and then the design itself, doing the design l
l 18 tradeoffs and the interactions with the regulatory authorities 19 so that these understandings and learnings can be factored l
20 into an economical and acceptable design.
21 (Slide) l l 22 HR. GRIFFITH: Now a little bit about the budget, 1
l 23 just to finish up--the cost of this is shown here. The green 1 24 is the facility costs at AHL west, mainly EBR 2, and the other l
() 25 costs are associated with the metal fuel development, safety HERITAGE REPORTING CORPORATION -- (202)628-4868
32 1 tests, and the pyroprocessing development, and these are about 2 70, 80 million dollars a year projected budgets.
3 And finally, where we have come from, where we are, 4 we can look back to the early days of the breeder program when 5 the LHR technology development costs were in the 800, 6 approaching a billion dollar range. We have gotten these down 7 now to where they are more in the $200 million range, and that 8 presently includes the cost of FFTF which may in the future, 9 may not be chargeable against this program, so we think we 10 have gotten the program down to a, a compact, focused program 11 that is affordable.
12 You can see that the budgets havs stabilized here 13 even with FFTF charges still costed against it, and we think
()
14 we have a good base to proceed to develop this LHR, and one 15 last time, it's absolutely necessary in this design activity 16 that early interactions and feedback with the regulatory 17 authorities be arranged and factored in, so again, I 18 appreciate the opportunity to talk to you today, and start
. 19 these understandings, continue these understandings.
! 20 DR. REMICK: On the slide, the expenditures are DOE j 21 expenditures only?
22 HR. GRIFFITH: Yes. These are government.
I 23 DR. REHICK: DOE?
24 HR. GRIFFITH: DOE, AEC, ERDA.
i
[} 25 DR. REHICK: But not NRC?
HERITAGE REPORTING CORPORATION -- (202)628-4888
33 1 HR. GRIFFITH: No.
2 CHAIRMAN WARD: So the LHR budget there, you know, 3 let's look at next year, is a little over $200 million. Is 4 that right? Am I reading that right?
5 HR. GRIFFITH: Yes.
6 CHAIRMAN WARD: And the IFR Argonne West total then
'. is 7 million. The previous chart for '09 says 70 million.
8 HR. GRIFFITH: The numbers are there. Let's--well, 9 for '89 it is 70. For '90, it is 86.
10 CHAIRMAN WARD: About a third; now where--I am 11 trying to get an idea of what other R&D work is going on and 12 will be going on over the time span that you have got shown >
13 there other than the IFR Argonne West work.
() 14 HR. GRIFFITH: The major additional funding would be 15 for the steam generator tests at ETFC, and they are like $16 16 million a year, and seismic testing and steam generator.
17 Seismic is only about half a million of that, though.
- 18 The other major items right now of R&D are being ,
19 worked by the industrial vender, GE, with attempts, attempt to '
20 align internationally with development of certain of the 21 needed components, and as an example, they are working with 22 Ensaldo in Italy to develop the fuel handling equipment. They ,
j 23 are looking to Japan to align, to do other part, and they will 1 24 do and complete that R&D on international money, and they are
(} 25 integrated into their program, so we have the operation of EBR i
HERITAGE REPORTING CORPORATION -- (202)62J-4888 l
l 34 )
l 1 2 in developing some control and fuel and what not
-w 2 understanding. !
x-) i 3 We have the other steam generator and component 4 testing that can be done at ETEC. By the way, within that IFR 5 technology is the EH pump work, and those are the principal 6 areas now, and I think that our ir. tent at DOE is to stay 7 compact and focused down in this range until we better 8 understand the design, and so we are doing still in 9 conceptual, advanced conceptual design stage, interacting with 10 NRC to establish whr.t the thing will look like, and then we 11 will start working on the more detailed component performance.
12 CHAIRMAN WARD: I guess what is bothering ne is that 13 the work you just described which seems to be the remaini*ag
() 14 two thirds of the budget, is mostly in, it scens to be very 15 necessary development to support, pretty directly support the 16 design.
17 I look at the previous chart, the safety analysis 18 and tests, and the core design' work are sort of dropping down i 19 to very little toward the end of the period you have shown 20 there which would seem to indicate that you feel you have all 21 those questions satisfactorily answered, and I'm, it just 22 seems to me--maybe we will talk more about this later in the 23 meeting--that there is an awfully big scale-up that has to 24 occur. I mean the EBR 2 core performance seems to be very
[} 25 well understood and there are analytical tools for predicting HERITAGE REPORTING CORPORATION -- (202)628-4880
35 1 it and there has been a fine series of tests to confirm those,
- ,- 2 and to help develop the analytical tools, but there is a big k/ scale-up from EBR 2 to PRISH or whatever the thing is going to l 3 t
I 4 be, and I don't see where the work to establish the adequacy 5 of those tools and the adequecy of the understanding of the 6 really pretty complicated core behavior that you are depending 7 on, you call it inherently safe, is really going to work in 8 the big core as well as it seems te work in the EBR 2, and I 9 don't, it is not clear to me where that work is.
10 HR. GRIFFITH: That's a challenging question. Let 11 me tell you what is here. $
12 The work that is described here is considered 13 adequate, is planned by Argonne to complete the referenced
() 14 fuel composition to be operated in EBR 2 and recycled through 15 the recycled plant into EBR 2, and at that time the 16 understanding then from the fuel tests that that system works.
17 That's what is described here.
18 There is additional work at FFTF in which metal fuel 19 driver, the metal fuel--for economic reasons, the department 20 decided to convert FFTF to metal fuel when the U-024 cores 21 available are accepted, so that's unified for production 22 capabilities.
23 As a result, FFTF will have reference link fuel 24 e.'ements in this. All that, the driver fuel is not the one l 25 for the ( mercial design, but they would have the capability O
l HERITAGE REPORTING CORPORATION -- (202)628-4888
36 1 of introducing reference fuel elements into that design.
~
2 Now I have to say right now that the use of FFTF is 3 being, is being looked at to use FFTF for plutonium 238 4 production for space and defense missions, and if that became 5 its principal mission, we would still be able to put fuel 6 elements in there and test it, but we would not get to a 7 larger fuel core test, and so the question is coming in now as 8 to how one would approach the ommercialization, and one might ,
9 have to do it with more emphasis in that area during the 10 demonstration plant phase, and we believe a demonstration 11 plant would be needed for this.
12 How--we scaled up from EBR 2 to FFTF the first time 13 on the oxide fuel, and while there clearly will be some
() 14 uncertainties, they could be worked out in the demonstration 15 plant.
16 CHAIRMAN WARD: Okay.
17 DR. REMICK: While we are on the DOE budget, is DOE 18 funding any research on or continuing funding research on the 19 core accidents?
20 HR. GRIFFITH: In the oxide area, no. In the metal 21 fuel area, there is some small efforts to look at the 22 phenomenology of such a thing happening, but we would like to 23 better an approach that guaranteed us we didn't get there with 24 an acceptable probability.
[} 25 DR. REMICK: Where is that work being funded?
HERITAGE REPORTING CORPORATION -- (202)628-4888
37 1 HR. GRIFFITH: For the phenomenology?
2 DR. REMICK: Yes.
3 HR. GRIFFITH: That's at Argonne.
4 DR. REMICK: And how large is that?
5 HR. GRIFFITH: I don't know the dollar value, but it 6 is not large.
7 DR. KERR: Mr. Griffith, can you tell me whether the 8 DOE goal is that the LHR be safer than water reactors, about 9 as safe, or none of the above? And in reading the NRC SER, I 10 find it is somewhat ambiguous, and--
it HR. GRIFFITH: I think we need interactions with NRC 12 on this matter. But let me, let me say--
13 DR. KERR: Is DOE not setting its own goal? Are you
() 14 going to let DOE set the goal for you?
15 HR. GRIFFITH: NRC has, has a policy statement and--
16 DR. KERR: I am trying to find out what DOE's goal 17 is if it has one.
18 HR. GRIFFITH: We think these reactors are much 19 safer than that. That's the way we would like to hear it--
20 DR. KERR: I'm sorry. My question was what is your 21 goal? Is your goal to make them safer, about as safe?
22 HR. GRIFFITH: Our goal is to make them safer.
23 DR. KERR And measured in terms of what?
24 HR. GRIFFITH: Heasured in terms of the probability
(} 25 of releasing the fission products to the public, measured in HERITAGE REPORTING CORPORATION -- (202)628-4888
38 1 terms of radiation exposure to the people who operate the
/~T 2 plants.
V 3 DR. KERR: And how much safer? I mean how much less 4 risk?
5 HR. GRIFFITH: Our goal is to make the plant so that 6 it is impossible to release the fission products. That's our 7 goal.
8 DR. KERR: I wish you hadn't told me that. I won't i 9 ask any further questions.
10 CHAIRMAN WARD: Any other questions? Thank you very 11 much, Mr. Griffith--very informative presentation.
12 Let's go right to Mr. King, who will be next.
13 HR. KING; My name is Tom King. I am with the 14 Office of Research, and my branch has the lead for reviewing 15 the three DOE advanced concepts.
- 16 What you are going to hear today is a summary of our l 17 review on the PRISH reactor. It is--a couple of months ago we r
18 presented to you our review and our SER on the HHTRG. The j 19 review process we followed on the PRISH and the SAFR that you l t
20 hear about for the next few months is essentially the sme as 21 we followed on the HHTRG. The conclusions are somewhat 22 different. as you will hear today.
i l 23 Basically the purpose of today's presentation is to ;
i
- 24 summarize our review and conclusions. We consider this an i
() 25 information briefing only. We are not asking for an ACRS I
f l
HERITAGE REPORTING CORPORATION -- (2021628-4888 i
39 1 letter at this time, although we would hope after your 2 November Full Committee to get a letter fr the Full 3 Committee.
4 I am going to cover the background and overview of 5 our review process, overall conclusions. Ralph Landry is l
- 6 going to take you through chapter by chapter the key points, 7 the key issues, that came out of the review and the key things 8 in our SER. John Flack will talk about the PRA, and Greg Van 9 Tule from Brookhaven National Laboratory will summarize the 10 independent analysis that we had done on the PRISH design.
! 11 CHAIRMAN WARD: You mentioned SAFR. So you do plan j
j 12 to go ahead as you had scheduled .11 along with the SAFR SER?
13 HR. KING: At this point in time, we have not
() 14 stopped any work on SAFR. We are proceeding to finalize that 15 SER, and plan to come to you unless our management decides q
16 otherwise.
17 (Slide) 18 HR. KING: The status of the PRISH SER, basically 19 the review is complete. The chapter, the verse of the SER we 20 sent to you back in the beginning of September does need some 21 updating, particularly those three chapters that I have listed 22 on the vugraph. We will provide the ACRS an updated copy in 23 time to support the November Full Committee meeting.
24 Hopefully by the end of this month sometime we will have an l
[} 25 updated copy.
HERITAGE REPORTING CORPORATION -- (202)628-4888
i 40 j
1 Basically we are hoping by the end of November to 3 2 have completed our internal reviews, including the NRR and CJ 3 CRGR. !
4 (Slide) 5 HR. KING: By way of background, our review started 6 back in December of '86. We have received from DOE back at l 7 that time two main documents--the preliminary safety 8 information document, and the probabalistic risk assessment
. 9 for the PRISM. We had a series of formal meetings to go ;
10 through those documents chapter by chapter. Those lasted from 11 April of '87 to November of '87. We had a couple of follow-up 12 meetings. As a result of those meetings, we issued a series 13 of questions. They resulted in a series of amendments. ACRS
() 14 I believe has copies of all anendments as well as the initial 15 submittals from the designers.
16 We briefed, the Subcommittee was briefed twice on 17 the PRISH design and the Ful3 Committee twice on the PRISM 18 design, and there was one meeting on EBR 2. You concentrated, ;
I 19 the Subcommittee concentrated on the metal fuel aspects of the ;
6 20 sodium reactors. We don't intend today to go back and give I i
21 you a complete design review, although in Ralph Landry's 22 presentation, he will hit and describe some of the key 23 features in the plant that relate to some of the issues and l 24 R&D itetas that we want to talk about. We are not going to go
(} 25 through and give a design description por se.
HERITAGE REPORTING CORPORATION -- (202)628-4888
41 1 (Slide) 2 MR. KING: The purpose of our review was to provide 3 quidance early in the design process on the acceptability of, 4 licensibility of the designs, which includes guidance on the 5 licensing criteria, potential designs to meet those criteric, i
6 and the acceptability of the R&D proposed by DOE to support 7 the design.
8 Basic guidance that the staff used in conducting the 9 review was that given in the Commission's advanced reactor 10 policy statement, which basically said let's perform early a
11 interactions with the designers. Let's recognize that these 12 plants use simplified and innovative means to accomplish their l J 13 safety functions and to review them on their own merits, and
() 14 the guidance of the staff in tarms of what would be an !
15 acceptable level of safety for these designs was that they !
16 should provide at least the same degree of protection in the 17 public and environment that is required for current generation 18 LWRs.
19 They went further to say that they did expect these 20 designs to provide enhanced safety, although it was not put 21 forth as a requirement in the Commission's policy statement. ,
22 DR. KERR: How is the safety provided by the current 23 generation of LWRs described? Becauce when you were comparing 24 it with that--
(} 25 HR. KING: There is no single measure of safety for HERITAGE REPORTING CO.iPORATION -- (202)628-4888
- 42 l
l 1 an LWR. I think it .s described by a lot of things, and it is I
l 2 a collective--
0 3 DR. KERR: I mean you are doing this review so you
)
4 must have something that you are comparing--
f 5 HR. KING: That is coming up on the next vugraph.
6 We will talk about how we did that.
7 DR. KERR: And it is your view that the goal of the 8 Commission is that these be as safe as current generation of 9 LWRs?
10 HR. KING: That's the requirement.
11 DR. KERR: Even though the advanced LWRs are 12 supposed to be or expected to be safer than the ctreent 13 generation?
34 HR. KING: By current generation LWRa, we have f) 15 defined that in our NUREG 1226 as being those advanced licht 16 water reactors that are currently under review by the 17 Commission, the ABWR.
18 DR. KERR: You don't mean the operating?
11 HR. KING: I don't mean the ones that are operating.
20 I mean the ones that are currently under review today that 21 will fall under the guidance provided on severe accidents by 22 the Commission's policy statement.
23 DR. KERR: We don't really know how safe those are P
24 because nobody has established that in fact we don't really 25 know quite how one is going to deal with a severe accident O
HERITAGL REPORTING CORPORATION -- (202)628-4888 L
43 1 review of those other than on an ad hoc basis?
(} 2 HR. KING? The staf f is formulating how to deal wi' 3 severe accidents on those. Consistent with that formulation, 4
i 4 we have proposed for the DOE plants that we are looking at a 5 treatment of severe accidents that we believe is consistent 6 with that, t
, 7 DR. KERR: So when you decide how to deal with the i
8 advanced LWRs, then you will know how ;'ou, how safe one j 9 expects these plants to be? Is that--
l l
10 HR. KING: Certainly when the LWRs, it is nailed 11 down how to treat them on LWRs, we are going to have to be
! 12 sure that what we proposed for these plants still is I
i 13 consistent with what was decided on the LWRs.
i 14 DR. KERR: Thank you.
l 15 (Slide) 16 HR. KING: The staff review process was laid out l
I j 17 originally back in the SECY paper at the end of December. It t
i 18 is also summarized in NUREG 1226, which includes a definition i 19 of the advanced reactors, includes a definition of current l
20 generation LWRs, and it summarizes our review approach in that f
21 we are going to build upon applicable LWR regulations and l 22 criteria in doing the review. We are going to utilize the l
23 Commission's policy statements, safety goals, severe accident l 24 standardization as guidance for us to develop criteria for 1
() 25 these designs, and that we are going to assess the supporting l
HERITAGE REPORTING CORPORATION -- (2021628-4883 l
-.- . ~ _ . . _ , _ _ . - . . . . . - - - . - - -,
44 1 R&D needs as well as the need for a prototype on these 2 advanced designs.
3 (Slide) 4 HR. KING: If you recall, we went on and before we 5 presented the SERs to you we presented two Commission papers 6 that tried to deal with what we felt were the key issues 7 associated with these designs.
8 One was standardization issues, which have now been 9 incorporated into the proposed rule on standardization, 10 CFR 10 Part 52, 11 And the second one was on the key technical issues, 12 the range of accidents which we felt needed to be considered 13 in these designs, the use of a mechanistic siting source term,
() 14 the containment of emergency planning questions. Those papers 15 are still before the commission for their guidance, but we 16 have completed the HHTRG review and we are completing the 17 PRISH and the SAFR reviews, using that, the staff 18 recommendations in those two SECY papers as the gd. dance for 19 the, completing the review and the documentation in ti.e SERs.
20 Basically what we did in reviewing the designs, we 21 compared the PRISH design to get to the question of how do we 22 measure equivalent level of protection does the plant have as 23 compared to an LWR 7 24 The way we approached that was we went through and
(} 25 compared the PRISH design to all those factors that we feel HERITAGE REPORTING CORPORATION -- (202)628-4888
I 45 1 contribute to safety on an LWR. For exemple, an LWR reactor 2 vessel is designed to certain ASMB section Class 1. We went 3 through and compared the proposed design of the P'. M reactor 4 vessel to see if it was equivalent, and where the PRISM 6 reactor vessel went into high temperature conditions, it was a 6 requirement that a high temperature code case be developed to 7 support that design.
i 8 We went through and looked at redundancy and 9 diversity in shutdown systems, decay heat removal systems, and 10 went through basically all of those factors it was, we felt
- 11 were important, and compared PRISM against them, where we 12 could use the same guidance for an LWR directly to PRISM. We 13 did that. Where something was different, we tried to come up
() 14 with some criteria that we felt contained an equivalent level i 15 of protection in that area. We didn't go in and measure it 16 using, comparing PRA results for PRISM againrt PRA results for l 17 LWRs. We didn't go ** and compare PRA results for PRISM 18 against the safety goals. It was more of a deterministic t
J i 19 review process. -
20 DR. KERR: From what I saw, that is an accurate 21 description of what you did, and I don't have any great 22 quarrel with that other than the fact I think the general
- 23 design criteria for light water reactors are probably
- 24 obsolescent, but in the standard review plan, if I remember
() 25 correctly, one does now have some quantitative reliability HERITAGE REPORTING CORPORATION -- (202)628-48S8
46 1 criteria for certain key systems in LL'-s. As far as I could 2 tell, you did not mention the possibility of using that O 3 approach to what one might consider to be key components of '
4 PRISH.
5 Are you likely to do that, or did you deliberately 6 decide not to do that?
7 HR. KING: The only, as I recall, the only thing the 8 standard review plan has in terms of reliability is in the 9 auxiliary feedwater system for PWRs.
10 DR. KERR: That's true, 11 HR. KING: These plants don't have an auxiliary 12 feedwater system.
13 DR. KERR: But I mean this was chosen not I think
() 14 because it was more system, but it was because it was 15 considered a very key component that would decrease reactor l 16 risk.
I 17 What I am saying is there is at least some precedent 18 in the current LWR review for picking key components and 19 suggesting or requiring that they have quantitative 20 reliability goals.
l 21 Did you deliberately decide aot to do that?
22 MR. KING: Yes. We deliberately tried to stay away 23 from putting down numerical reliability goals that it would 24 then have to be rigorously demonstrated that they are met. At 25 this stage of the review, we tried to stay away from that.
O HERITAGE REPORTING CORPORATION -- (202)628-4888
47 1 DR. KERR: Does that mean that you are unlikely to 2 introduce them later, or have you not decided?
3 MR. KING: No. I think if at a later date we would 4 find that such goals would be useful, and such goals, there is 5 a way to demonatrate they can be met, that is acceptable to 6 the staff. We considered that. I am not going to promise 7 anything, but we certainly considered that. At this stage, we 8 didn't, we didn't take that approach.
9 In the review we tried to concentrate on those areas 10 that we felt were key to the viability of this design, 11 particularly the safety attributes that were being relied upon 12 to demonstrate that these plants achieve a high degree of [
13 safety.
() 14 In doing that, we ended up selecting what we call a 15 set of bounding events that were selected by engineering 16 judgment that we felt would bound the uncertainties in system 17 performance and system reliabilities and things that were 18 proposed in the PRA. Again, we didn't select the range of 19 accidents strictly on a probabalistic or reliability type 20 considerations. We certainly looked at the PRA, considered 21 that, but we wanted to put a sufficient challenge on this [
22 plant through some deterministic engineering judgment that we 23 felt the safety argument would be fairly clear that this was a ,
24 safe plant.
l
(} 25 We didn't--one of the other fundamental things we HERITAGE REPORTING CORPORATION -- (202)628-4888
48 1 were shooting for was these designs were intended as proposed 2 to us to prevent core melt, to prevent significant release of 3 fission products. Therefore, in looking at the design 4 criteria what we, ultimate criteria that wo looked for was do S these designs in fact keep the plant out of core melt 6 conditions? For the bounding events, do these plants keep you 7 out of a sodium boiling condition?
8 We felt if we got into those conditions, well, we 9 are really into an area where we didn't know very much about 10 how the plant would behave and we would be very hard to, to 11 make an argument that the containment system or the .tvacuation 12 plans and so forth would be sufficient.
13 As you recall, back on our criteria we proposed for
() 14 the containment issue to the Commission, one of the, one of 15 the items in there was that for us to accept a design without 16 a conventional containment, the design had to eliminate core 17 melt accidents, positive reactivity accident, positive 18 reactivity feedback accidents from the list of events that had 19 to be considered in the design.
20 DR. KERR: Mr. King, I am reminded of the early 21 1960s and the question of possible core melt in light water 22 reactors, transition was made from smaller to larger reactors, 23 and a committee was appointed to investigate the possibility 24 of coolinga badly damaged core, and the committee concluded
(} 25 that you really couldn't predict coolability of badly damaged HERITAGE REPORTING CORPORATION -- (202)628-4888
49 1 cores, and the thing to do was to make certain that they 2 didn't happen, that this didn't happen, and this relates to 4
O 3 the emergency core cooling systems, and the evaluation models 4 and all of these things. And after that, for a long time it 5 was concluded that cores couldn't melt in light water 6 reactors. It took us a long time to realize that they could.
7 I hope we are not making that same mistake in this 8 analysis to say well, we don't know how to deal with positive 9 core coefficients so we are going make sure they don't happen,
- 10 that positive events don't occur.
I 11 It is not a conflict. It is just something about 12 which I am uneasy.
i 13 MR. KING: I think no matter what design you have in
() 14 front of you, you can always put enough its on the accident 15 situation to pretty much make anything happen you want to have 16 happen.
l 17 These designs have taken great pains to try and l
) 18 limit the likelihood of a core melt or of a positive i i l 19 reactivity feedback accident. In reviewing the design, we t
20 wanted to see is there, was the design sufficiently, was there ;
L 21 sufficient test put on the design to bound the uncertainties 22 that we felt still existed that we could reach the same !
23 conclusion, that for these very low, unlikely events, you l ?
l 24 still don't reach core melt conditions, or sodium boiling :
i !
j 25 conditions. Certainly you can go beyond that and assume, you l
[}
t l HERITAGE REPORTING CORPORATION -- (202)628-488F L
50 t
1 1 know, both vessel and guard vessel failing and all that l 2 sodium, and all these have a problem, and I think what we are O 3 trying to say is we tried to define how you described a range ,
i 4 of accidents for which we felt this plant had to be analyzed
- 5 for. ;
6 We have also looked beyond that in terms of are 7 there any cliffs? We have done some looking at energetics.
8 We have done some looking at core melt, in-vessel coolability 9 from the standpoint of what kind of margin do we have beyond 10 that range of accidents on, that we are laying on the l 11 designers.
1 I
12 We feel that to accept a design without the 13 conventional containment building, we have to have a pretty l 14 high confidence that these bad accidents aren't going to
(]) l l 15 happen, and we have proposed a set of criteria that should
} 16 make that decision, and what we have done is bounced the 17 design off those criteria. The next slide I am going to give i
- 18 you a summary. ;
i 1 19 HR. AVERY: I'm not sure I understand the bottom I line you reached. You say you looked at some aet of--
20 !
21 CHAIRMAN WARD: Bob, would you swing the microphone f l
l 22 in front of you?
4 23 HR. AVERY: You looked at subset of bounding events.
24 Did any of the bounding events that you looked at lead to core 25 disruption or not?
i O .
I l HERITAGE REPORTING CORPORATION -- (202)628-4888
[
51 1 HR. KING: Yes.
{} 2 HR. AVERY: Some did?
3 HR. KING: Yes. We have a problem with the design 4 which you will hear about.
5 DR. SIESS: You said that if you do accept a design 6 without a containment, you have to have a higher confidence of 7 no core melt than if it did have a containment.
8 How much higher? Is the containment good for a 9 factor of ten so if you take the containment off, you have got 10 to have an order of magnitude lower probability of core melt, 11 or containment give you a factor of a hundred?
12 MR. KING: Essentially--
13 DR. SIESS: A containment is not perfect. You can't
( 14 say I have a containment. I will accept a very high 15 probability of core melt because a containment will keep 16 everything in. Or will you say higher level of safety without l
17 a containment, does that reflect in some way your confidence 18 in the ability of containaant to contain it? Do you have any 19 idea what order of magnitude? Two, or factor of two or what?
20 HR. KING: The criteria we proposed to the 21 Commission in SECY 88-203 basically said we want to look at a 22 range of events for these plants that wauld assure we wouldn't 23 have a core melt or positive reactivity feedback accident down s 24 to a cumulative probability of ten to the minus 6th.
() 25 DR. SIESS: Total?
HERITAGE REPORTING CORPORATION -- (202)628-4888
52 l
1 HR. KING: Total.
2 DR. SIESS: If you have a containment, would you 3 raise that ten to the minus 6th to ten to the minus 5 or two 4 times ten to the minus 6 or what? I mean it would be 5 different, wouldn't it?
6 MR. KING: It could be different.
7 DR. SIESS: You said that if there was one way it 8 ought to go to the other. You said without a containment, you 9 have got to have a lower probability of core melt, so if 10 somebody volunteered to put a containment on, you ought to be 11 able to give them back some of that.
12 HR. KING: The yardstick we used, a ten to the minus 13 6, we said we didn't want to exceed the Part 100 dose limits
() 14 at the site boundary at ten to the minus 6th.
15 DR. SIESS: If I put a containment on--
16 MR. KING: We felt if a plant without a containment 17 had a core melt or a positive reactivity feedback accident, we i
18 couldn't guarantee that we would meet that criteria. If you 19 put a containment on, how much can you back off in terms of a l
20 core melt probability? We didn't piopose a number in our 21 criteria because what was, the designs, we were being asked to 22 look at these designs from the standard point of what does it i 23 take to eliminate the containment? If you were to put a 24 number in there, you can back off some.
[} 25 DR. SIESS: You are not prepared to say how much?
HERITAGE REPORTING CORPORATION -- (202)628-4888
1 53 1 MR. KING: I'm not prepared to say how much. I l 2 think in the upcoming safety goal paper that the staff is 3 probably going to send to the Commission shortly you will ses 4 some guidance in that area.
5 DR. SIESS: I have been hearing those words for 6 three years.
7 DR. KERR: Holp my memory. My impression is that 8 the goal that the French used in Phenix was core melt 9 frequency ten to the minus 6 with containment. Is that--
r 10 MR. KING: That doesn't--
t 11 DR. KERR: That was what I was thinking, but I'm not 12 sure. Okay.
13 HR. KING: We have to look into that.
() 14 DR. SIESS: They talked about ten to the minus 6, 15 but they don't want to demonstrate it.
16 DR. KERR: They have some goals, and one of the
> 17 consequences was the shutdown system they devised which was i
18 highly reliable, redundant, diverse, and I remember your SER 19 said you did make use of your experience that had been gained ;
20 with Phenix and Superphenix, so I thought perhaps you did this 21 in comparison? i 22 MR. KING: I don't--
23 DR. KERR: Continue.
\
24 MR. KING: We have lool:ed at their numbers, their !
l
(} 25 critr a ria.
54 i 1 To summarize the conclusions of the SER, first l
{} 2 one--keep in mind this is a pre-application SER. This was '
3 preliminary guidance we are talking about. It is not a design J
4 approval. It is not our intent in this SER to resolve all ;
5 issues. At this stage we would expect a large number of 6 issues and open items that still remain in the SER. !
i 7 Our intent is to get those documented, get those on .
8 the table so that this design effort proceeds and the staff ;
< i 9 review follow-up efforts proceed. We have in front of us the f
10 list of items that need to be worked on. I i t 11 In summary, on the PRISH design, our overall ,
j 12 conclusion is that there are a couple of fundamental design i i 13 issues which need to be resolved before the staff can conclude j
( 14 that the PRISH design has the potential to achieve a level of l 15 safety with the current generation LWRs. These involve the 16 response to certain bounding events which lead to sodium
! 17 boiling or lead to a core melt condition, that the response of i 18 the plant to those accidents tie directly to the acceptability 4 I i !
, 19 of the containment design proposed for PRISH. I i
l 20 You will hear more about those particular events and [
21 the response of the plant, what our independent analysis said f l 22 as we go through the presentation today. We will also go !
s >
'3 through the complete list of bounding events so you can see i I l 24 what they are. Basically there are four of them that gave us j l
() 25 trouble--flow and blockage of a single assembly, loss of power f t
HERITAGE REPORTING CORPORATION -- (202)628-4888
l l
55 l
l
! I to an EH pump, transient overpower ATWS event, and the loss of I
2 the safety grade decay heat removal system for a period of l 3 time.
l l 4 We think there is a potential for design solution 5 for all of these, although at this stage we are not going to 6 hold up the SER waiting to resolve what the acceptable design
> 7 solution would be for those.
8 (Slide) 9 HR. KING: In general we feel--
10 CHAIRMAN WARD: The SER does indicate what sort of 11 design solutions might be possible or necessary?
12 HR. KING: We can put some examples of potential 13 design solutions. We don't want to dictate anything to the
() 14 design. We think it is really his job to go back and decido 15 how you best want to address these issues. We will certainly 16 discuss some of the things. If we put anything in the SER at 17 all, it will just be in terms of examples, not in terms of 18 requirements.
19 DR. KERR: Mr. King, in 1-4, which is tied to the 20 safety issues, there is the sodium void, and the description 21 seys there is a question as to a void coefficient.
22 Does that mean whether the void coefficient will be 23 positive or how big it will be er what am I to conclude from 24 that?
(} 25 HR. KING: There is a positive void coefficient on HERITAGE REPORTING CORPORATION -- (202)628-4888
56 t
4 1 the plant. The issue is we get into sodium boiling under
{} 2 certain situations that leads to positive reactivity feedback i 3 accidorst. That's the issue. ,
)
~
4 DR. KERR: The issue is not you're prepared to
! 5 accept a positive void coefficient. It is just a question of l
6 whether the void can occur, so it is not, it is not a void j
i 7 coefficient issue. It is a void issue.
- 8 HR. KING: It is a void issue. There is events
) 9 where the void can occur, and then you are in trouble with the
) 10 cause of the void coefficient.
i 11 DR. KERR: I would have some misgivings about that j 12 also.
13 So are you going to require that it not be a
, () 14 positive void coefficient, or that there be a demonstration of I
3 15 that voids can't occur?
j 16 MR. KING: It is the latter that we are looking l
i 17 toward now, although I think in our discussion with the i
! 18 designers, one approach they may want to try and take to
}
i 19 resolve the problem is to try and reduce the void coefficient.
i
, 20 DR. KERR: Thank you, J
21 HR. KING: Beyond the fundamental issue of i
l 22 containment question, we feel that the final l
23 determination--there is also other issues that you find in the 24 SER that you hear about today. We feel the final
{
I
() 25 determination on the PRISH acceptability is continued
(
HERITAGE REPORTING CORPORATION -- (2021628-4888
57 [
1 basically upon resolution of all the issues in the SER. ,
l 2 completion of the analysis and R&D programs that the designers (
3 and DOE have proposed, and Commission action on the proposed l i ;
j 4 criteria for the four key issues.
i t 5 Certainly final determination of PRISM acceptability l
6 involves completing the normal licensing process that the f i
{ 7 staff reviewed, and it also is contingent upon the successful
, 8 design, construction, operation and tenting of prototype i l
- 9 reactor. The PRISH design prototype was proposed. It was I
! 10 part of the DOE proposal. They did need a prototype to (
) i 11 support that design, so we don't have an issue there, but it l
? l
) 12 is an item that we agreed needs to be done. !
j [
13 CHAIRMAN WARD: Well, what, as far as the second !
14 itom there, the lictasing review, th're are, I mean with your i l 15 SER, and also with the informal discussions you had with the }
l i 16 DOE people which seem to have been, you know, at icast part of f I t I
17 the base of their decision, do you think there is a, enough of !
18 an Agency commitnent to certain positions or acceptance of l
1 19 certain design features that the licensing reviewers ten years
)
! 20 from now are going to have some problems and be constrained by I 1 21 positions that have been already taken? [
i ;
I '2 HR. KING: The whole purpose of this review is to !
l I put some early guidance on the street from the Commission. f l 24 Now that's one of the reasons we wrote those two SECY papers l 25 to the Commission and raised the policy issues to them is to i !
l
)
! HERITAGE REPORTING CORPORATION -- (202)628-4888 f f f
58 l
1 get their views out on how to treat some of these issues.
, 2 Also send the SERs to the Commission before they are published C O 3 so that they can see what the staff is proposing. ;
- 4 We hope with that kind of endorsement by the L
5 Commission, that this guidance we put out or, t s street will
! 6 have some weight, and will be followed by the staff ten years j i
l 7 from now. If one of these thinge comes in, that guidance I
.8 won't just be a research report. It will be an Agency report t j 9 that was reviewed and endorsed by the Commission. If there is l
) 10 a new Commission in ten years, I can't predict what is going f I
11 to happen in ten years. The intent is to try and put as firm i
3 12 a guidance on the street today as we can. !
i i
{ 13 CHAIRHAN WARD: So you think that reviewers in the i
14 future will be constrained by the guidance that you are 15 providing today? !
l 1
I
! 16 HR. KING: We hope the guidance we are providing ,
3 i I 17 today will lay the ground work for them tc continue the review l t
18 and they will build upon it. If there is new information, j l
19 something we didn't think of that cones up, certainly it isn't l 20 going to constrain them from putting that on the table, but we !
I i i
21 hope it lays the ground work so that they are not starting [
" from a clean sheet of paper again and going back and repeating l I L
) 23 all the work we have dona over the past two years. I l f j 24 CHAIRHAN WARD: But thtt might hold for a research j l !
25 document just as well as for a, an SER' And what I am trying [
l O ;
i !
HERITAGE REPORTING CORPORATION -- (202)628-4888 i i
59 l l
1 to figure out is, you know, to what extent are the decisions l
2 you made, both the guidance you have given and the guidance !
O 3 you haven't given, going to establish, be established as the i
4 Agency's position on these issues up, you know, in the l
5 licensing review? !
6 HR. KING: It is preliminary guidance. The SERn do 7 not have any legal standing in terms of setti'g regulations or I
8 being used in any way in a licensing hearing rpplication. The i
9 staff will havs to produce its normal SER and go through its l I
10 formal review process on an actual application. When the ;
i 11 staff starts thet review, we are hoping that they have some 12 foundation to start with in terms of knowing what the issues f I
13 are, what the R&D needs are, and hopefully we will h&ve, we t
O 14 witt a v ia atiri a t 1t twe x r i o aa ed et=tiea-15 to those key issues so they are not starting from ground zero ;
16 and as well as the designers aren't starting from ground zero.
l ll 17 They will know what the r:ommission's views are on the key [
18 issues, t 19 CHAIRMAN WARD: Okay. Thank you. j l 20 (Slide) .
21 HR. KING: The last vugraph--I don't want to leave j 22 anybody with the impression that PRISH design is bad. I think i
l, 23 the PRISH design has a lot of favorable characteristics. All l !
! 24 three of the designs have a lot of favorable characteristics.
j 25 It has the potential for only minor core damage or no core
- i i !
HERITAGE REPORTING CORFORATION -- (202)628-4888 ,
)
l 60 l 1 damage or Vission product release over a wide range of severe l
2 challenges to the plant, does, is a design with reduced f l (
3 dependence on human actions and reduced vulnerability to human j i [
l 4 error, operator error. !
5 The long response time on the reactor under many 6 accident conditions is a good feature, providing time for 7 evaluation and corrective action, and we believe that design, j 8 design such as this that are attempting to prevent core j 9 damage, significant radiation release under the severe 10 challenges, does provide an advantage in that you can test the 11 reactor under those conditions, demonstrate that that does in i
12 fact happen, and that would be one of the purposes of the 13 prototype test, to confirm the response in the plant, surface 14 charaisteristics of the plant to some of these severe
(])
15 challenges, so we think it does have a lot of potential. What l
t 16 you are going to hear about today are the issues, design and 17 R&D issues, but I want you to keep in mind we fool the design .
6 18 does have a lot of potential.
19 CHAIRHAN WARD: Okay. Thank you very much.
20 DR. KERRt Mr. King, on the page 1-5 of the SER, the 21 statement is eado that low probability events beyond the I
22 traditional design basis envelope are considered in the design !
I 23 so as to provide a sufficient challenge to the plant to allow f 24 the use of a mechanistic siting source term. j 25 I didn't understand what that meant. Can you tell O i HERITAGE REPORTING CORPORATION -- (202)628-4888
l 61 l 1 me what--
2 HR. KING: Light water reactors, when you look at
/}
3 siting the plant, you use Part 100 dose guidelines and you use !
4 the TID 14844 source term, which is a source term that is i i
5 supposed to be representative of a ccro melt accident.
[
t 6 Now we are looking at designs that have gone to f
- / great lengths to try and elimincte core melt accidents, f 8 eliminate significant radiation release, so what we are f l
9 proposing instead of laying on them a large arbitrary source j 10 term, which is representative of some core melt, we are trying 11 to give them credit for the enhanced safety features that they i
, t l 12 have put in their design and the, what we propose to do is to t l
, 13 have them analyze their response of their plant to a range of l 1
( 14 what we consider very severe accidents, very severe challenges I
15 I should say, and if their design does in fact limit radiation i 1
i l j 16 release under that not of severe challenges, then we are ;
a i
j 17 proposing that the siting source term be the source term that i l
- 18 comes out from those severe cha11anges, not some arbitrary j i i l 19 source term that we put on the designers. !
\
I {
l 20 DR. KERRt One might describe this as a design basis c 1 !
! 21 source term? !
l l
22 HR. KING: Siting source term is what we call it. }
r
! 23 HR. AVERY: What would you propose for the source !
l
! 24 term? Let's say for the events that they look at, the severe l I
() 25 challenges. fuel does not fail. What then would--would there f t
i 1
HERITAGE REPORTING CORPORATION -- (202)628-4888 ,
I l '
i 62 1 be a source term? And if so, where would it come from? ;
2 HR. KING: We think it could be something maybe from 3 a radioactive sodium spill. Certainly some of the fuel pins 4 are predicted to release their fission gas. It could come 5 from a fission gas release through a relief valve. It could 6 come from a fuel handling accident. We would want to look at J
7 the range of those, see which one turns out to be the worst.
9 HR. AVERY: You would not impose that under those j 9 conditions, an arbitrary source term?
i 10 HR. KING: What we are saying is we would not then
) 11 put on them a big source term that would be representative of 12 a core melt accident like we do on light water reactors.
J 13 DR. RERR: In 3-1, I find it is important to note
() 14 that the proposed criteria allow tradeoff between plant i 15 protection and accident mitigation to achieve equivalent level
! 16 of safety. Then it says they do not allow elimination of 17 either plant protection or mitigation.
18 Since the only mitigation system I know of in LWRs l
i i 19 is conventional containment. since apparently conventional
- 20 containment is going to be eliminated, where is the i 21 mitigation?
1 22 HR, KING: WELL, there is a containment vessel. It f
23 also doubles as the guard vessel around the PRISH reactor.
j 24 DR. KERR: It was like saying there is a vessel as 25 far as I am concerned. That's hardly--
[}
HERITAGE REPORTING CORPORATION -- (202)628-4888
63 1 HR. KING: It is underground in a silo which will
{} 2 provide some mitigation in terms of fallout of fission 3 products, se we feel there are some mitigative features.
4 True, it is not a conventional neat tight containment 5 building.
6 DR. KERR: So there is the vestiges of mitigation?
7 That is typical mitigation one might call it. Okay. Thank 8 you.
9 HR. KING: Is that it?
10 DR. KERR: Mr. King, I also want to compliment you 11 on those overheads that you used today. They are legible and 12 readable, and that was helpful.
13 HR. Ki'NG: We received a strong message to not come
() 14 down with the usual.
15 DR. KERR: Even more important, you responded.
16 CHAIRMAN WARD: Okay. Thank you very much, Tom.
17 Let's take a break until 10:25, and then Mr. Landry will pick 18 up.
19 (A brief recess was taken.)
20 CHAIRMAN WARD: Let's pick up, and we have Hr. Ralph 21 Landry.
22 HR. LANDRY: Thank you. My name is Ralph Landry. I 23 am the project manager for the Office of Research for both our 24 PRISH and *.he SAFR preliminary design.
() 25 As was mentioned earlier, there in a fair amount of HERITAGE REPORTING CORPORATION -- (202)628-4888
- - - =_-. - . - - ...- -
64 ;
4
[
1 liquid metal reactor experience worldwide at the present team.
1 2 It is not as extensive as the experience which we had with the
[}
- 3 light water reactors because there is quite a bit of I
4 experience. l
! k g 5 DR. REMICK: Could you tell me which one of those r
> 6 reactors have ordinary containment?
f 7 HR. LANDRY: The EBR 2 reactor in Idaho has a steel !
f d
8 shell around it, that while it looks like a conventional 1 t j 9 containment, it doesn't have the strength of a reinforced [
i 10 concrete containment with steel liner.
i 11 The Phenix reactor has what is more or less a t
. 12 conventional containment. It is not the full high pressure
- 13 containment, but it is a containment building.
() 14 PFR, I'm not familiar with the type of containment l
l 15 that has.
16 FFTF I believe has a full containment building j 17 around it. The NR 300 hasn't been built and operated yet.
j 18 MONJU is under construction, and Superphenix in France has a f 19 full containment structure. The containment structure is not i
20 exactly th9 same as we are accustomed to with the light water 21 reactors. The Superphenix plant has a double wall reactor i 22 vessel. containment vessel, as we have had at, for PRISH and 23 SAFR, but it has a machine tome built sbove the reactor 24 vessel, containment vessel, and open upper head region. That
)
() 25 machine dome provides a containment function off che reactor HERITAGE REPORTING CORPORATION -- (202)628-4888
65 1 vessel itself. Around that machine dome is another building 2 which is a strong concrete building. It is not the full high O
I 3 pressure containment which we are accustomed to in PWRs, but l 4 it is a reinforced concrete containment.
l 5 DR. REMICK: The German and Japanese, do you know 6 what the intent is there?
7 HR. LANDRY: No, I don't. I am not familiar enough ,
l 8 with those plants. I know the French changed with the plant 1 9 for Superphenix 2 or what is called sometimes Superphenix 2, 10 sometimes called R&N 1500. That plant would do away with the l
11 machine dome. It would do away with a the reinforced concrete !
12 building and would simply be a semi-leak tight confinement 13 building above the reactor. It would retain the double wall
() 14 vessel, though, and it would retain the very thick, very 15 strong upper head closure. That plant is much larger. The 16 Superphenix is a 1200 megawatt plant. The R&N 1500, as the 17 name implies, would be 1500 megawatt electric plant.
18 DR. REMICK: Thank you.
19 CHAIRMAN WARD: Ralph, could you begin with that?
20 Just a r;inute. The LWR containments have a fairly 21 straightforward basis for design. Is there some--you 22 mentioned several of these have a containment of some sort, 23 and in some cases I guess it kind of looks like an LWR 24 containment. It is heavy concrete and maybe it is even dome
{} 25 shaped. I don't know. But is there a consistent basis for HERITAGE REPORTING CORPORATION -- (202)628-4838
t 66 1 the designs of the containments that we see here?
2 HR. LANDRY: With the LWR or the PWR in particular, 3 the containments are designed to withstand the design basis !
l 4 losn of coolant accident pressurization and temperature. The l 5 containments which have been designed for much of this work l 6 because this system is low pressure and low energy fluid, by 7 comparison to the pressurized water reactor, are designed to {
l 8 withstand the loadings from sodium fires, or disarming of (
9 radioactive material, rather than withstand the high pressure, 10 high temperature exposure that they would get from a loss of :
I 11 coolant for a PWR. [
12 CHAIRHAN WARD: So for all of them you say the
(
1 13 sodium fire is the basis?
14 HR. LANDRY: I can't say that for sure for all of 15 them. I do know that for some, the sodium fire is considered E I
16 the maximum loading, particularly for temperature, j 17 CHAIRHAN WARD: Thank you. f 18 (Slide) 19 HR. LANDRY: I don't want to spend a great deal of 20 time on the PRISH design, but just to refresh everyone's f i
21 memory, the PRISH design is, as we reviewed it, it is intended 22 to be nine reactor modules. Reactor modules would form a 23 nuclear island which would be the maximum protection, 24 safeguarded island. Three reactor modules would be connected 25 to their three steam generators which would be, as Jerry HERITAGE REPORTING CORPORATION -- (202)628-4888
67 !
1 Griffith said earlier this morning, industrial grade and ;
2 outside the r.uclear island. These three modules would then be I O 3 connected to--
I 4 DR. KERR: Excuse me. Do you know what Jerry f i
5 Griffith means by industrial grade? !
i 6 HR LANDRY: Again, I shouldn't say that. We f 7 interpreted industrial grade to mean less than N stamp, that !
8 there would be high quality construction, high quality !
i 9 material, but would not go through the pedigree and quality !
10 assurance necessary to receive an N stamp.
11 Primarily the difference between N stamp and other i
12 high, very high quality industrial material is the pedigree, ,
l 13 and the following through of the material, follawing through
(
() 14 on the quality control of the construction.
I 15 CHAIRMAN WARD: Is there evidence in the existing j 16 population of plants that the N stamp imparts higher ;
17 reliability than let's say good industrial grade? ;
l 18 HR. LANDRY: I don't believe there is evidence to i l
19 the contrary. It is not--to answer it by coming at the back !
20 end, it simply means that with the N stamp materials, we have i i
V 21 a good track of the quality control. Vith the non-N stamp, we 1 e
22 don't have the same tracking. Therefore, with any kind of ;
i 23 failure, or faulting, N stamp materials, we have a paper 24 trail. With the non-N starp, we don't have that same quality j 25 of paper trail. That doesn't mean that they are less quality.
(]) l l
HERITAGE REPORTING CORPORATION -- (202)628-4888
68 1 It simply means we can't prove it. :
Y 2 M9. WYLIE: Well, there is a more exacting quality 3 control f .m on that N stamp material; well, from the source 4 all the way through.
5 HR. LANDRY: And we review and approve that.
6 HR. WYLIE: From the regulation standpoint that is 7 true. I mean from an actual quality standpoint, I don't think [
8 there is any argument that you are getting a better product 9 in the end.
I 10 CHAIRMAN WARD: Well, it is just that in several 11 attempts to, you know, to confirm that assumption, in you f
L 12 know, reliability studies, let's say associated with risk [
13 analysis, PRA work, I haven't, there have been some research [
() 14. attempts I guess or research attempts to confirm that, that j 15 assumption, and I haven't maybe there is something that has !
}
i 16 been done, but I have never seen any that really has confirmed i i !
I 17 it. !
- I 18 HR. WYLIE: I don't know about mathemati s of f I t
19 proving it, but from an actual practical usage standpoint, I ,
I 20 think that-- i l 21 CHAIRHAN WARD: Yeo are convinced that it does? l
. t
?
l' 22 HR. WYLIE: Oh, sure. Thern are too many things of
{
- I l 23 the manufacturer of the material, discarding, you know - t
! 24 that's engineering, t
25 DR. REMICK: Help me in that vugraph. I looked at 1 i l t I
HERITAGE REPORTING CORPORATION -- (202)628-4888 i l
1 69 1 the reactor facility to be, it appears to me it is all below
]
l 2 grade. There is no building above it at all. The only
- 3 building above ground is the RVACS stacks. There is no 4 building above grade level?
5 MR. LANDRY: I will show you a slide of that.
6 (Slide) 7 HR. LANDRY: The design as we have reviewed it would 8 place the entire reactor module, the upper head area, the 1
9 equipment vaults, instrumentation vaults, the ducting of the l
10 tunnel containing the steam line or excuse me, tne i 11 intermediate sodium line, all below grade. The only portion i
12 that would be above grade would be the RVACS stacks, i
j 13 DR. REMICK: There is emple room there for refueling
() 14 and so forth? The top of the module is at grade level?
15 HR. LANDRY: The design that has been proposed would l
i 16 incorporate ample room for refueling.
i
) 17 (Slide) l
.l 18 HR. LANDRY: Unfortunately, these drawings aren't 19 that good. The intent would be if thr*e was on-site 20 reprocessing--this, of course, we did not review, but simply i
- 21 looking at on-site fuel cycle, the reactor service building i
22 would incorporate train tracks and a specially designed 23 refueling module which would be moved over the top of the 24 upper head access area.
I 25 The plate that is in place would be removed. and the
[}
l l HERITAGE REPORTING CORPORATION -- (202)628-4888 l
70 1 refueling machine would lower a cannister over the top of the
{} 2 head and perform the refueling operation. All of the 3 refueling would be performed from above this area, and the 4 design is to incorporate a sufficient room to move that 5 machinery and/or train tracks and out on train tracks.
6 DR. REMICK: One would be doing that changing 7 internals or anything like that? I guess I just find it 8 surprising there is no building above there.
9 MR. LANDRY: There is room. This is about the best 10 drawing that I have. There is supposed to be sufficient room 11 in this upper head access area to permit any equipment i
12 changeouts that are necessary that do not incorporate the use 13 of the refueling machinery.
) 14 DR. REMICK: That's what I was wondering about. I 15 realize it is just conceptual. I was wondering if there were 16 provision for cranes and so forth to do that. Apparently you 17 are saying that is the intent? .
18 MR. LANDRY: Well, what we have looked at to date,
! 19 there are no cranes in that area. Crane operations could be l
! 20 by the refueling machine or another crane could be brought in.
l l 21 We haven't discussed that in detail with the designer.
! 22 CHAIRMAN WARD: What is the dimension here and what l
l 23 is the length of the cover? Just sort of give us a rough l
24 idea.
()
25 DR. REMICK: Core is four feet I believe.
i l
l l HERITAGE REPORTING CORPORATION -- (202)628-4888 l
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71 1 MR. LANDRY: The core is 47 inches. I am trying to g- 2 see if I have a diagram that shows dimensions on it. No, I L) 3 don't. The vessel, reactor vessel, is 19 feet across.
4 CHAIRMAN WARD: Okay.
5 MR. LANDRY: The length of the vessel would be 60 6 feet and the core would be down in this region and would be 47 7 inches of active fuel.
8 MR. LANDRY: I have got everything out of order 9 here.
10 The overall plant parameters for the plant design 11 which we reviewed call for nine reactor modules producing a 12 total of approximately 1200 megawatts of electricity. They 13 would he ganged in groups of three, three reactors and 400
() 14 megawatts thermal leach, and here you see the sodium flow 15 rates, the sodium temperatures. There would be one control 16 room to control all nine modules. The plan as we have 17 reviewed would have one operator controlling each gang of 18 three modules. The cycle for the steam would be saturated.
19 DR. REMICK: Has the staff addressed that yet?
l l 20 MR. LANDRY: Yes. You will hear about that later.
i 21 DR. REMICK: All right. I assume the staff has 22 caught the typo in the actual draft SER. That indicates that 23 the intermediate sodium outlet temperature is 41,000.
I 24 MR. LANDRY: It is not quite that hight That
{} 25 temperature is actually only around 800, 875 degrees. That i
}
HERITAGE REPORTING CORPORATION -- (202)628-4888
i 72 ,
1 was intended to be the sodium, the secondary sodium flow rate.
2 (Slide) 3 MR. LANDRY: I think we would have a little trouble 4 with boiling sodium if it got that hiphi 5 The design review which we performed was based on 6 use of the four policy statements--advanced reactor policy 7 statement, severe accident policy, the safety goal policy 8 statement, and the standardization policy statement.
9 The factors which we considered in the review were 10 such items as are conservative engineering design practices 11 used in this design? Is there a redundancy and diversity in 12 the design? And is there a valance between deterministic and 13 probabalistic analyses? And finally, is there defense
() 14 in-depth?
15 DR. KERR: Can you tell me why after years of living i
16 with the single failure criteria, and recognition that really 17 it doesn't set any particular standard of reliability, we are j 18 still using it? I mean now we could use reliability criteria.
i 19 HR. LANDRY: Well, single failure criteria we are 20 familiar with, and they give us a conservative approach to 2i design, what we have traditionally called design basis events.
22 It p,1ves you an added--
23 DR. KERR: Wait a minute. Why does that give you a 24 conservative approach to anything?
() 25 HR. LANDRY: It adds one more failure on to the HERITAGE REPORTING CORPORATION -- (202)628-4888 L
73 1 failure assumed that it creates--
2 DR. KERR: That's conservative only if a single 3 failure criteria provides you with enough reliability.
4 What I am saying is, it says single failure 5 criterion, does not give you any known or designated level of 6 quantitative reliability, so I don't feel there is anything 7 conservative about the failure criteria.
8 HR. KING: We preserved it in our review to maintain 9 its equivalency in looking at this plant versus an LWR, and we 10 preserved it in what we call the design basis accident region.
11 Beyond that, we are looking at multiple failures.
12 DR. KERR: I recognize that you preserved it. I'm l 13 curious as to why you did. After all these years of doing
() 14 work on reliability, and having some feel for what reliability 15 is achievable, it would seem to me it would make some sense to i
(6 specify reliability rather than talk about single failure 17 criteria.
i 18 HR. KING: Again, I don't think we are prepared to 19 state our criteria in terms of reliability. It is going to 20 have to demonstrate that those things are met. I think we are 21 doing it in a little more traditional way of specifying 22 redundancy, single failure, multiple failure, ways of 23 achieving reliability without putting down a specific number 24 that you have to show that you have met.
(}
25 CHAIRMAN WARD: Well, I thought, you know, earlier HERITAGE REPORTING CORPORATION -- (202)628-4888
. , _ - . _ _ ~ _ _ _ . _ _ _ - _ _ _
74 1 you said you didn't want to dictate design features.
2 HR. KING: For the issues that are, we have got in
(
3 the SER, we are not telling them how to solve those issues.
4 CHAIRMAN WARD: But when what you are really 5 interested in is reliability, by insisting on the single 6 failure criter.a, you are in fact dictating a design feature.
7 I point out I don't think you are being consistent, which is, 8 you know, not unusual for any of us.
9 HR. LANDRY: We will take your comment into 10 consideration. The site which is proposed, has been proposed 11 for the PRISM plant references the Gessar 2 site. The only
. 12 exception taken to the Gessar two site is the snow load that 13 is considered for the PRISM design. This is primarily because
) 14 it is a below grade design so it can withstand a higher snow 15 load 16 CHAIRMAN WARD: It isn't so much that you perceive
- 17 the need for it to accommodate a larger snow load? It just i
18 looked like it was easy to do?
19 MR. LANDRY: It has the capability. Why not just 20 take credit for it?
21 CHAIRMAN WARD: Okay.
22 MR. LANDRY: That said, the proposed siting criteria 23 for PRISM would envelope 90 percent of the current reactor 24 sites. In other words, there is no proposal for a
() 25 metropolitan site. The siting would be matched up with the HERITAGE REPORTING CORPORATION -- (202)628-4888
75 1 current concept of a reactor site.
'N 2 DR. REMICK: I don't want to follow that statement.
)
3 That does not preclude a metropolitan site, would it?
4 MR. LANDRY: The Department of Energy has not 5 proposed a metropolitan site. They have only proposed that 6 the siting criteria encompass 90 percent of what are the 7 current licensed sites.
8 DR. REMICK: Right. I guess I interpret your saying 9 that precluded because there are no metropolitan sites now. I 10 don't come to the same conclusion.
- 11 MR. LANDRY
- That's an extension of what I said. .
12 What we .:.can is that the plant which we have reviewed does not 13 propose metropolitan siting. We have proposed--
(/ 14 DR. REMICK: Doesn't propose e.ny siting?
15 MR. LANDRY: They have not proposed a specific site.
16 They have proposed siting criteria which would encompass or 17 envelope 90 percent of the current sites. We did not review 18 it on the basis of any site different than those siting 19 criteria.
20 DR. REMICK: I apparently misunderstood what you 21 said.
22 CHAIRMAN WARD: I still don't understand. What sort 23 of criteria are you talking about?
24 MR. LANDRY: Such items as the hydrology, seismology
() 25 and in particular, the demographic distribution.
HERITAGE REPORTING CORPORATION -- (202)628-4888
76
( 1 CHAIRMAN WARD: Demographic?
2 HR. LANDRY: There will not be demographic
}
3 distribution worse than the 90 percent of the current sites.
4 DR. KERR: Would you permit site, encourage siting 5 one of these plants near a large body of fresh water?
6 HR. LANDRY: There is that possibility. One of the 7 proposals for the siting would be that this site could be a 8 barge shipable site, which of course could be a large body of 9 fresh water, with the proposal to perform extensive 10 fabrication of the facility in a factory, that the proposal 11 would then be to ship by whatever means are available--rail, 12 or barge shipable.
13 Briefly, the earthquake phenomena which are cited
() 14 for this plant are, of course, zero period acceleration of 15 .3G, and in a few minutes when we talk about the reactor 16 module itself, we will talk about seismic isolation in module, 17 the spectrum which were used in defining the ZPA from the Reg 18 Guide 1.60.
19 DR. KERR: This barge would have to be sunk below 20 the lake level to achieve the underground location?
21 CHAIRMAN WARD: That's just the barge for transport.
22 HR. LANDRY: The barge would be for transportation 23 of the hardware.
24 CHAIRMA!! KERR: I see. So you are siting near a 25 large body of water, simply pass over the body of water, but
(])
HERITAGE REPORTING CORPORATION -- (202)628-4888
i 77 l
1 you might move inlan after that?
(~ 2 MR. LANDRY: We have not considered siting of the
(
3 facility on a barge in the water, no.
4 DR. KERR: Okay.
5 MR. LANDRY: If that's the question.
6 DR. KERR: Well, I was thinking of there are some 7 locations which--I think of two immediately, one in the Great 8 Lakes. Would you encourage that?
9 MR. LANDRY: We are not encouraging any particular 10 site. We have only reviewed the siting criteria and said that 11 the siting criteria which have been proposed have been 12 determined to be acceptable.
13 DR. KERR: But would the NRC regularly staff likely
( 14 preclude such siting?
15 HR. LANDRY: We would review the site for the 16 specific plant as it was submitted, and based on the 17 enveloping characteristics of a plant, the siting 18 characteristics of the proposed site, and our requirements at 19 that time, review the site and determine if the plant could be 20 cited there.
21 HR. KING: We didn't include in the scope of our 22 review specific sites, or these questions that you are 23 raising.
24 DR. KERR: You said something about hydrology.
() 25 MR. KING: We looked at the site to see if it HERITAGE REPORTING CORPORATION -- (202)628-4888
78 1 complied with the Reg Guide 4.7 requirements for siting, which
(} 2 covers a lot of things. Beyond that, we didn't entertain any 3 peculiarities of siting of this plant. That was not part of 4 the review.
5 (Slide) 6 HR. LANDRY: Earlier we mentioned that we considered 7 defense in-depth in the review. Four of the factors which we 8 look at in our concept of defense in-depth are prevention, 9 protection, mitigation, and emergency plannina.
10 For prevention, we looked to see that the sort of 11 conservative design assumptions had been made. We looked to 12 see that the appropriate codes and standards had been applied 13 in the design description, and we looked to see if there is
) 14 quality in the design, and construction and planned 15 maintenance of the plant.
16 At this stage, we have not reviewed the quality 17 assurance program for such items as design, construction, 18 maintenance, and operation. However, we will do that should 19 the design continue further and an application come in.
20 In the area of protection, we looked to see that the 21 systems necessary for protection of the plant are designed 22 with the adequate reliability and incorporate the redundancy 23 and diversity.
24 In the area of mitigation, we have looked to see
() 25 that should a fission product release occur, that there are HERITAGE REPORTING CORPORATION -- (202)628-4888
I l
79 j i
1 systems to delay or limit the release of those fission
{} 2 products, and we also looked to see that--
3 DR. KERR: You are talking about a protection 4 system. Is it your view that if the protection particularly 5 on the reactivity control system, were shut down, is it as 6 reliable here as it is on say something like Phenix, or did 7 that enter into your review at all?
8 MR. LANDRY: I am not familiar enough with tha 9 shutdown system at Phenix to answer that.
10 DR. KERR: It puzzles me a little that you are 11 taking into account experience gained with other liquid metal 12 cooled reactors, you did not look at the Phenix. At least 13 that there has been a great deal of experience.
() 14 MR. LANDRY: We have only--with Phenix, there has 15 not been adverse experience, but we have not reviewed the 16 actual reliability data.
17 MR. KING: We looked at it from the standpoint of 18 Clinch River and CRBR shutdown systems. Design principles and 19 reliability should be equivalent to those two design systems.
20 DR. KERR: Clinch and CRBR have never been operated.
21 MR. KING: FFTF has.
22 DR. KERR: Phenix has. I mean it just seems to me 23 you have listed these as background, I assume with experience 24 available, and it would certainly seem to me that one would
() 25 want to look at the reliability claim for the shutdown system.
HERITAGE REPORTING CORPORATION -- (202)628-4888
80 1 You may decide it is not what you need, but at least it would
~h 2 make some sense to understand what it is and why it was chosen (O
3 to be what it is.
4 HR. LANDRY: At this stage, we have not obtained the 5 reliability data on all of those plants that were listed as 6 the operating experience base. They are out there, and we 7 have given a cursory look .o the operating experience at a 8 number of those plants and we have seen that the operating 9 experience has been favorable. There have not been serious l
10 problems. This is only a conceptual design phase, and the l 11 next phase, of course, we would expect to go into a great deal 12 more detail.
13 MR. KING: Getting to your question, we didn't just
( 14 look at the design and say yes, that's a shutdown system. We 15 looked at it to see where did it come from? Is the d(42.gn one 16 that builds upon the Clinch River and FFTF experience? It is 17 not some new wildly different design that nobody has ever seen 18 before.
19 DR. KERR: But the French have at least two separate 20 shutdown systems, if not a third. They may need one much 21 worse than you do, but it would seem to me that it would be 22 worth your while to find out why they chose this approach, ,
23 what their criteria were, just because they have had a lot of {
l 24 experience, they have given it a good deal of thought, and I l 25 mean apparently it would seem to me that the reliability of
(:)
HERITAGE REPORTING CORPORATION -- (202)628-4888
l 81 1 the system being proposed here is not as good as the French 2 system. Maybe it doesn't need to be.
O(>
3 I think the assumption is you don't really care 4 whether you should--I shouldn't put it that way, but it isn't 5 as important that you get this reactor shut down. That may be 6 true, but it sure seems to me it is nice to be able to control 7 criticality of a reactor fairly reliable, and I see no 8 evidence of any quantitative criteria being used. Perhaps 9 they can't be used. If you make the shutdown unreliability 10 something like ten to the minus 8, then you probably can't 11 achieve that, but I don't know what it is that one thinks 12 should be achieved in this case.
13 All I see is a single failure criteria and
(} 14 redundancy and diversity, and it seems to me we, we ought to 15 be beyond that phase with all the work that has been done 16 around here on reliability over the past ten, fifteen years.
17 MR. LANDRY: We will consider those comments when we t
l 18 get into the next phase. You will be hearing, though, later 19 today our probabalistic risk assessment which was performed, j 20 and the assumption on reliability under shutdown systems are 21 presented in that document:
1 22 DR. REMICK: A question I should have asked on the 23 previous slide--I assume we have heard about all we are going 24 to hear about site?
25 MR. LANDRY: All I was planning.
O HERITAGE REPORTING CORPORATION -- (202)628-4888 1
82 1 DR. REMICK: In the SER, on your Section 2 site
(} 2 characterization, there is a statement that kind of puzzled 3 me. It says that the assumed site has excluded the following 4 hazards, excluded aircraft impact and explosion, and I can 5 understand that site specific, but yet you chose to observe 6 that it had an increased snow load capability. -
7 It seems to me that maybe the design is not far ;
8 enough along that you could also say that it could withstand 9 explosion at a certain amount of force or perhaps it could 10 withstand the impact of aircraft at this particular site or 11 something. I was wondering why you specifically excluded 12 those? Or maybe you don't have enough information, but you 3
13 made an observation on snow load, so you must have something
( 14 on structures.
15 HR. LANDRY: We just made the observation that the 16 site which was proposed was the Gessar 2 site with the or.ly 17 exception that the designer has proposed to increase the snow 18 load over the Gessar 2 site.
19 PR. REMICK: It is not because that is the structure 20 he is proposing?
21 MR. LANDRY: No. This is the proposal from the 22 designer.
i j 23 DR. REMICK: Also you excluded fires. What kind of l
l 24 fires were you excluding there? Were these forest fires, 1
() 25 internal type thinigs?
l l HERITAGE REPORTING CORPORATION -- (202)628-4888 l
83 1 HR. LANDRY: Yes. In the Chapter 3 of the 2 preliminary safety information document, the designer has 3 chosen to apply the general design criteria as found in 10 CFR 4 50, Appendix A, pretty well directly across the board to the 5 PRISH design. -
6 After lookina at what they have proposed, we, the ;
e 7 staff agreed with the designer that 35 of tho.se GDCs are t
8 directly applicable to the LHR. They are not spr.cific to a !
F 9 light water reactor. We agree with the designer in the 10 exception that they have taken to six of the GDCs, and we i agree with them that there are nine GDCs which are not 11 12 applicable because they are so unique to LWR designs.
13 However, the staff has noted that there are five I
() 14 GDCs which we do intend to modify in the future for the next i
15 round. We feel that they can be made applicable, and simply 7 16 because the wording today which applies to an LWR does not 17 preclude their utilization for other designs, we do intend to 18 modify them and make them applicable to the advanced designs.
19 (Slide) 20 HR. LANDRY: As Tom mentioned earlier, we have !
21 reviewed the design basis accidents and beyond design basis 22 accidents which have been proposes. We have reviewed the PRA 3 i :
c 23 which has been submitted, and we have noted after that review 24 that because of the uncertainties in the PRA, we feel that we 1
l
(} 25 need other events beyond the design basis and beyond design 1
HERITAGE REPORTING CORPORATION -- (202)628-4888
- l
84 1 basis events which will bound those events. i 2 We have proposed a series of bounding events which 3 we expect to provide conservatism in selecting a site-specific l 4 source term. These events have been broken down to cover such ;
i .
5 categories as reactivity excursions, loss of off-site power, ;
6 loss of forced cooling, steam generator tube rupture, the 7 large sodium leak, and the external events. We have not 8 defined external events at this time. We will define those in !
l 9 a later stage of the design when we have that definition also l 10 for the LWRs. We intend to make the definition consistent ;
.; 11 with that imposed on the LWRs. ,
12 Now in this list, we have, we have looked at i 13 inadvertent withdrawal of all control rods during in their
() 14 most reactive range for a number of conditions. One is !
l 15 without SCRAM for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. We have looked at that condition l
j 16 then with imposing only forced cooling or imposing no forced L 17 cooling and only emergency shutdown cooling. i 18 We chose the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> as you have heard in the past I 19 when we discussed the issues which are before the Commission, f I !
j 30 based on the requirement that the operator have at least 12 ,
a i
21 hours in which to recognize that an event is underway, and ,
!, l c
22 second, that there be at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in which to put in
] f I t 23 place an ad hoc emergency planning program.
(
At least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, this means that gives !
i 24 oR. xERR:
() 25 at least two shifts an opportunity to recognize that sonething (
HERITAGE REPORTING CORPORATION -- (202)628-4888
85 l 4
1 is wrong I assume? I 2 HR. LANDRY: Well, we didn't--we selected it, it O 3 does give two shifts, yA but it also recognizes the fact that 4 most of these transients that we have seen analyzed for the ;
5 liquid metal t eactors or for all the advanced reactors are 6 very long and slow in taking place, so we felt that we need a 7 longer period of time for the operator to recognize that he is j 8 coing into an event which is going to get him into trouble.
9 CHAIRMAN WARD: Let's see. This inadvertent 10 withdrawal has presumed some rate of withdrawal?
11 HR. LANDRY: It is drawn at the maximum rate .
12 possible.
13 CHAIRMAN WARD: Possible under what circumstances?
(} 14 HR. LANDRY: By the control system for the reactor.
15 CHAIRMAN WARD: So there is some reliably, reliable 16 on, ebsolute limit on withdrawal rate inherent in the control 17 system?
18 HR. LANDRY: If the control, if the motors were--the 19 withdrawal the rods have a stepping speed of certain number of 20 inches per second. We assume that they have withdrawal at 21 twice that. This is a limitation on the withdrawal rate.
22 That's the limitation we assume.
f 23 CHAIRHAN WARD: Because it would take nore than a 24 single failure to--
25 HR. LANDRY: Because the motors on this have been O
HERITAGE REPORTING CORPORATION -- (202)628-4888
86 1 designed so that they cannot withdraw at a faster rate than 3 assumed. The control circuitry and the power signatory for 3 the design are such that sure, the motor is trying to exceed 4 the maximum withdrawal rate, that would trip the circuit 5 breaker.
6 MR. XING: Multiple failure events here on all of 7 these; these ari not single failure events. To go through the 8 rationale, take the first one, why did we pick that as a 9 bounding event? We picked it because what was presented to us i
10 was the design. It was the designer said I don't think my 11 control room is a safety function anymore. Doesn't matter 12 what goes wrong in that control room. I designed my plant to 13 be able to handle it, handle it in from the standpoint of not
() 14 having any significant core damage or release of radiation.
15 We have said all right, let's test that. Let's 16 assume there is a fire in the control room, and it shorts or.t 17 your control system such that the worst possible thing cou?.d 18 happen. All the controls rods come out and you have got hTWS 19 on top of that. And can your plant take that for 36 hou'ts?
20 That's how we came with up with the bounding event. The 36 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> comes from the emergency planning criteria that de put 22 out. We were trying to be consistent with that; that the 23 design also said we don't want to have off-site emergency 24 evacuation, or pre-plant drills, early notification. Rememb2r
[} 25 the criteria we proposed to the commission said if you can HERITAGE REPORTING CORPORATION -- ( 202 ) 6'J 8 - 4 8 8 8
87 1 show you don't have significant release for the first 36 2 hours, then we will accept that, so we have tied all of these 3 into 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Tied it into that criteria. It is that kind 4 of approach.
5 We have looked at designs and said what is the worse 6 thing that can happen? You are telling me the design can 7 handle all of these things. Lct's put some bounding events on 8 it, not try and wave our hands down in the low probability 9 range because I don't think there is an upper limit 10 uncertainty, so what, some of these probabilities of some of 11 these events are to make a case, they are not going to happen 12 much.
13 DR. KERR: At what point do you start counting for
() 14 the, to get the 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />? Thirty-sir, hours after what?
15 HR. KING: Initiation of the event.
16 DR. KERR: I would have thought that 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> might 17 have been chosen because it had something to do with the decay 18 of fission products after say a core damaging accident, 19 followed by shutdown.
20 HR. KING: Thirty-six hours is based upon emergency 21 planning criteria. It says without pre-planned evacuation, 22 carly notification, it takes about we estimate based on past 23 experience, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is adequate time to move people on an ad 24 hoc basis. Take that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, add to it another 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for
(} 25 the operator to recognize what is going on, and try and HERITAGE REPORTING CORPORATION -- (202)628-4888
88 1 correct the situation.
2 DR. KERR: It is based on that?
3 HR. KING: Yes.
4 DR. KERR: Thank you.
5 CHAIRMAN WARD: You have looked, for example, at ,
6 equipment failures, for example, caused by control room fire? ;
7 HR. KING: Yes.
8 CHAIRMAN WARD: But what about looking at 4
9 inappropriate actions that the operator might take because he 10 is confused or strong-headed or something? Is there any--I 11 mean what inherent protection is there against defeating 12 certain control functions or you know, normal, normal 13 equipment parameters, for example?
() 14 HR. KING: For that first one, I think it would 15 bound anything the operator could do. Whether h3 pulls all l 16 the control rods out or a fire causes it to happen, it is the 17 same event.
18 CHAIRHAN WARD: I mean for example, you said if the 19 rods are going to dry, to drive out at a speed, you know, 20 faster than the, that they will trip out, but is there a way I 21 that the operator can defeat that trip? l 22 HR. LANDRY: No. No, because it is a physical 23 dependency on the electrical circuit. The operator can't j 24 defeat the circuit limits.
() 25 HR. KING: We have already assumed an ATWS event in l HERITAGE REPORTING CORPORATION -- (202)628-4888 1 - . _ -
4
- 89 1 there. Even if he defeats it, what we have assumed remains is 2 the inherent passive protection features of the plant, the 3 decay heat removal system, the negative feedbacks.
4 CHAIRMAN WARD: But the negative feedbacks, maybe
, 5 they are not very sensitive, but I mean what if the rod 6 withdrawal is at twice what you believe to be the maximum 7 speed or at ten times what you believe to be the maximum 8 speed? Are the inherent feedbacks still operative or not?
9 HR. LANDRY: We haven't analyzed that condition, 10 two, three times the design maximum speed. When we get into 11 the prototype test program, one of the parts of that program
- 12 will, of course, be testing all the systems. We will have had 13 time, of course. We won't have to see that. Of course, you i
() 14 cannot exceed the maximum withdrawal rate of the rods, t
15 CHAIRMAN WARD: Well, of course the point I am 16 making is one we have made before, of course, that you know, 17 our experience, the most severe power reactor accidents we 18 have had have been associated not with fires in the control s
l 19 room, but with operators doing a clever combination of things
, 20 that's the wrong thing to do, and when we have a question f 21 about how inherently resistant this design is, for example, to !
22 that sort of thing, and you--I realize it is a very difficult i
23 issue, but it's the issue.
! 24 HR. LANDRY: That has been pointed out to us by the l
(} 25 director of NRR also, that his major concern is making reactor i
HERITAGE REPORTING CORPORATION -- (202)628-4888
_ ___ , _ _ __ . - _ . _ _ . __ _ _ . _ _ . _ _ _ _ _ - - ~
90 1 idiot proof. We are not really sure of how to incorporate )
r' 2 that in reviewing the design other than the designer has said 1
3 the design is sufficiently strong and has inherent 4 characteristics which will preclude human errors, or will, 5 will preclude human errors getting the reactor into an 6 accident condition.
7 When we look at these bounding events, we did not !
8 give credit for other action, events started and proceeded.
9 There was no credit for operator action, but at the same time, 10 we did not say that the other--
11 CHAIRMAN WARD: Credit or debit?
12 MR. LANDRY: Did not say the operator did anything 13 wrong or did wrong actions because looking at these, we could
() 14 not go through and determine every possible action the 15 operator could take which could be good or deprivation to the 16 system.
17 MR. KING: I think in general if you look at these 18 events, there is not a whole lot the operator can do to--he 19 can look at the general sense, but that's it. True, we didn't 20 give him credit for doing anything good. We added one, you 21 will see on the next page we added one that is strictly human 22 error event--blockage of assembly, fabrication model, because 23 of our concern about human error coming in not only from the 24 kind of control room, it can come in from, if something
(} 25 happened on the fabrication side, that could cause a problen.
HERITAGE REPORTING CORPORATION -- (202)628-4888
91 l 1 HR. CARROLL: An extension of this is subotage by 2 the knowledgeable insider. How have you looked at that? ,
3 HR. LANDRY: We have reviewed the sabotage plan. We 4 have had our cafeguards personnel review it for both outside 5 ard inside sabotage, and they have made a number of comments.
6 The details of the sabotage protection plan, we cannot discuss 7 in open session.
8 HR. CARROLL: I understand. That is a good test of 9 these sort of things, to play the knowledgeable insider game, 10 and see what he could do given a number of hours to do it in.
11 HR. LANDRY: Not operate the number of hours, but 12 the extended period of time to alter signatures in such 13 matters; that has been reviewed extensively by our safeguards
() 14 people, and they have made a number of comments in that 15 regard.
16 MR. CARROLL: Can you share with us at least any 17 general conclusions?
18 HR. LANDRY: I will get to the safeguards section 19 later in the discussion.
20 HR. CARROLL: Okay.
21 HR. LANDRY: If you would like to wait and bring it 22 up at that point?
23 CHAIRMAN WARD: Is that okay?
24 HR. LANDRY: That is chapter 13.
[} 25 DR. KERR: Mr. Landry, why is this referred to as a HERITAGE REPORTING CORPORATION -- (202)628-4888
92 1 site specific source term? Is this--it would seem to me it is 2 a reactor specific source term that we are talking about 3 rather than site specific.
4 HR. LANDRY: It really is a reactor specific source 5 term, but it is used for the evaluation of the site.
6 DR. KERR: Okay.
7 HR. LANDRY: You are right. It is a reactor 8 specific source term.
9 DR. KERR: Thank you.
10 HR. LANDRY: We also looked at the, in addition to 11 the withdrawal of the control rods, we looked at a Station 12 Blackout. Can the reactor take a loss of off-site power, loss 13 of all power for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />?
() 14 We also looked at the loss of forced cooling, and in 15 particular, we were concerned here with loss of not only the 16 forced cooling of the reactor, but blockage of the emergency 17 shutdown cooling system, and in this case, that is the RVACS 18 system. We looked at the case where the system is locked 100 19 percent for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. After that point, there is a 25 percent 20 unblockage permitted. Operator action of some nature can get 21 in there and unblock 25 percent of that air flow to the 22 reactor vessel or to containment vessel.
23 HR. AVERY: Is that with or without SCRAM?
24 HR. LANDRY: That was with SCRAM. We also looked at 25 the instantaneous loss of flow from one of the four EH pumps HERITAGE REPORTING CORPORATION -- (202)628-4888
93 1 with coastdown of the other three pumps, and we also imposed
(} 2 looking at steam generator tube rupture.
3 Early in the discussion, we had specified a number 4 of steam generator tubes to, assumed to be ruptured, but since 5 that point we have looked at the experience that they had at 6 PFR in Great Britain, and the number of tubes that ruptured in 7 the event they had, and determined that we would allow the 8 designer to select a number of tubes to rupture in the steam 9 generator, but they must justify the number of tubes, must 10 have support for the number of tubes that they assumed to 11 rupture, and if they don't rupture all at one time, justify 12 the sequence in which the tubes rupture.
13 DR. SIESS: Excuse me. By assuming instantaneous i 14 loss of flow from only one pump, you are in affect ruling out l 15 common mode failures?
16 HR. LATIDRY: Yes.
j 17 DR. SIESS: On what basis? Is there, do you have a l
18 deterministic, have you got some sort of scena that leads to l
I 19 one pump failure but not to others, or is this just arbitrary?
20 HR. LAllDRY: That it is not completely arbitrary.
21 The--I don't have the, I don't have a good diagram. The PRISH 22 design is using four electromagnetic pumps, four EH pumps.
23 flow we are aware that EH pumps, once they lose their supply of I 24 electricity, stop instantly. It is not like centrifugal pump
() 25 that is going to coast down. The designer has put on a 1
HERITAGE REPORTIllG CORPORATIO!! -- (202)628-4888
94 1 synchronous machine, or a design number for each of the EM 2 pumps, dedicated to each pump, so should there be a loss of
}
3 power, that generator would run down the pump rather than 4 allow the pump to stop instantly.
5 The event which we have imposed is that--
6 CHAIRMAN WARD: That generator doesn't have a 7 separate power supply, but just provides mechanical inertia?
8 HR. LANDRY: It is through--the cable comes into the 9 EH pump, supplying power to the pump, but there is also a 10 tapoff to the synchronous motor. When, if the synchronous 11 motor senses that, a break in the current flow by the power to 12 the EH pump, that motor driven by flywheel will drive down, 13 coast down, and supply power to the EH pump. That power
() 14 supply to the EH pump would be pre-programmed by the designer 15 flywheel. It would match a flow coastdown of a centrifugal 16 pump so the EH pump would not suddenly see a loss of power.
17 It would see power supplied by that in such a manner as to 18 represent the flow that a centrifugal pump would experience.
19 CHAIRMAN WARD: Thank you.
20 HR. LANDRY: We pose the loss of one of two >
21 synchronous machines.
22 DR. SIESS: What kind of mechanism would you lose 23 one of those synchronous machines?
24 HR. LANDRY: We just said it is loss. It is just f
() 25 arbitrary.
HERITAGE REPORTING CORPORATION -- (202)628-4888
95 1 DR. SIESS: Why don't--couldn't you equally 2 arbitrarily say two of them?
3 MR. LANDRY: Because the synchronous machines are 4 separated physically. We stipulated that they must be 5 isolated and they must supply protected power systems to the 6 EM pumps.
7 DR. SIESS: So they are completely, the failures 8 would be completely random?
9 MR. LANDRY: We have assumed that they would be 10 random.
11 DR. SIESS: I forget the proper word. They are not 12 related?
MR. LANDRY:
13 We have assumed it is not common mode
() 14 failure of all four pumps or synchronous m& chines at one time.
15 DR. KERR: They can't be shut off from the control 16 room?
- 17 MR. LANDRY
- No. If the power is shut off to them, l 4 18 they drive the pumps down. '
l 19 DR. KERR: Unless you have a breaker between those 20 machines and EH pumps? !
) 21 MR. LANDRY: We have stipulated no breakers. They 22 are hard wired.
23 MR. CARROLL: Can the reactor be operated with less !
24 than four EM pumps? i
() 25 MR. LANDRY: I have to ask the derigner that.
HERITAGE REPORTING CORPORATION -- (202)628-4888 !
1 96 l l
l 1 DR. GYOREY: No.
I 2 HR. LANDRY: Dr. Gyorey says no. i 1
3 DR. SIESS: Why not? It is physically impossible. j 1
4 MR. LANDRY: I would have to go back and check the 5 controls that they have described to us. My recollection is j l
6 that the control circuits does not permit pulling rods unless l
7 all four pumps are operating. Unless the four EM pumps are i
8 operating, the operator cannot begin pulling the control rods.
9 HR. CARROLL: Supposing he is at full power and 10 losing it?
11 HR. LANDRY: That's the event we are looking at.
12 MR. KING: SCRAM on low flow; have low loss of 13 off-site power. All four pumps coast down, and the plant is
() 14 designed to handle that. What we are looking at is an event 15 where you instantaneously lose one, plant SCRAMS, the other 16 three coast down.
17 MR. CARROLL: I was just trying to get at is it a 18 possible condition where you only start with three operating?
4 19 MR. LANDRY: !!o . The indications we have is that, 20 are that the plant cannet be started on three pumps.
21 DR. REMICK: Looking at the steam generator tube 22 rupture, does this include looking at the potential effects on 23 the primary sodium system from possible pressure pulse going 24 back to the intermediate heat exchanger?
25 MR. LANDRY: That is correct. That's why we have F7RITAGE REPORTING CORPORATIO" -- (202)628-4888
97 l
1 said failure to isolate or dump the water from the steam
{} 2 generator, so that we enhance, not only see the pressure i 3 pulse, but we enhance it.
4 HR. CARROLL: When you discussed Station Blackout, ,
5 you said loss of all power on site.
6 Is there any important DC systems involved with 7 this, with this?
8 HR. LANDRY: It does have 1E class DC system.
9 HR. CARROLL: And you lose that, too?
10 HR. LANDRY: No. ,
11 HR. KING: Station Blackout is loss of all AC power. !
12 HR. CARROLL: That's what it normally means. He 13 said loss of all power off-site.
( 14 DR. KERR: Th3re is not class 1E emergency on-site 15 power available, is there?
! 16 HR. LANDRY: At this time, there is not a design l
l 17 including an on-site 1E turbine or generator system.
18 HR. KING: Class 1E battery, DC power.
l f 19 HR. LANDRY: But not AC.
l 20 (Slide) 21 HR. LANDRY: We have also imposed a boJnding event i
l 22 to look at the large sodium leak. We have looked at both the l t
i 23 critical size in the reactor vassel, not simultaneous rupture i I
24 of the reactor vessel and containment vessel, but simply the ,
() 25 reactor vessel, and we have looked at maximum moderate enotoy [
l HERITAGE REPORTING CORPORATION -- (202)628-4888
99 1 fluence system, break of the intermediate transfer system 2 piping O 3 We were aware of the discussion at one of the MHTRG 4 presentations where the Subcommittee was concerned about 5 imposing double guillotine ruptures on low energy pipes. We 6 at first considered douole guillotine rupture of the LHT !
7 piping, but after discussing this with our own mechanical 8 designers, our own materials people, we were convinced that 9 the instantaneous double guillotine rupture is, simply cannot j i
10 occur in t'
- s kine >
>ing, so we backed off and said we 11 would accept the modorr. energy system break size. l i l 12 As I said earlier, for external events, we have not 13 defined external events at this point. We will define them L
(} 14 when they are defined for the LWRs and make the external 15 events consistent with the LWR. l 16 Since this list was put together, and since the !
. I 17 draft of the SER was put together, which you have reviewed, we i e i 18 have added another event, and that is flow blockage of one !
19 fuel assembly. ,
20 After looking at events that could occur, we went .
21 back and felt that there is the possibility of a human error (
22 in either putting an oriface plate on the inlet or blocking :
23 the inlet or blocking the outlet of a fuel assembly, and that ;
1 I
24 fuel assembly getting all the way through the quality control
- 25 program, and getting inserted into the reactor.
I HERITAGE REPORTING CORPORATION -- (202)628-4888 ;
L
i 99 1 What we are looking at primarily here is not a 2 blockage due to material within the reactor, but blockage due O 3 to human error which puts in a pre-block assembly, and then 1
4 there is the attempt to start the reactor with the assembly.
l 5 We have not defined this.
I 6 CHAIRMAN WARD: There is not individual i
t 7 assembly--let's see--individual assembly flow monitoring?
l 8 HR. LANDRY: No, there is not.
I 9 CHAIRHAN WARD: All right.
[
10 HR. LANDRY: Okay. We are concerned about this l t
11 event due to the positive nodium density coefficient. We felt .
12 that this is not an event which we could rule out from ,
13 happening. This is not a human error event which we could
() 14 rule out. This hss been--
l 15 CHAIRMAN WARD: Is there individual assembly you {
16 have for temperature monitoring? .
17 MR. LANDRY: Not of every assembly, no. Therre are -
i
~
18 only six.
19 HR. AVERY: What would you be assuming in this I 20 event, that you detected early in the game and SCRAM, or you ,
21 would not be detected for a while? {
22 MR. L NDRY: We haven't defined this event :
t 23 rigorously at this point. We have just informed the d$ signers 24 of both PRISH and SAFR that they will be expected at the next j 25 design phase to analyze this event. We have not even asked
[} ,
i I
HERITAGE REPORTING CORPORATION -- (202)628-4888 L i
4 100 1 that event be analyzed at this phase. We want to close out 2 the review at this phase in the design. The next phase we ,
3 will look, preliminary feeling is we will look at the j
4 possibility of a blockage of the inlet, and/or a blockage of 5 the ottlet of an assembly. determine if there is a difference 4 6 in the two.
7 We would be interested in methods by which this !
8 could be designed out. Can a detection mechanism, monitoring c
9 mechanism be designed into the system which would prevent or 1 i
) 10 detect a blockage and shut the reactor down before you ended I
i 11 up with severe fuel damage condition? .
I i
12 HR. AVERY: You are more concerned with the void i
13 reactivity than the potential for propagation?
l () 14 HR. LANDRY: We tre concerned with, our real concern i
- 15 is with the propagation. Con we void, will the void cause i
16 melting, slumping or resolution of the fuel in that assembly l
17 which would cause a reactivity insertion sufficient to cause ;
i
! 18 propagation of failure to the adjoining assemblies around the j
19 core.
i
- 20 HR. CARROLL
- Given the attention that potential l
l 21 accidents under refueling conditions has been getting on the i
22 PWR business, the mid-loop problem end so forth, have you 23 looked at, in detail at potential accidents in the refueling i l
24 mode? {
(} 25 HR. LANDRY: Those have been looked at within the i
- HERITAGE REPORTING CORPORATION -- (202)628-4888 L
101 1 design basis cccident, and envelope of that. We have not gone
(} 2 beyond the accideets which have been analyzed by the vender.
3 HR. KING: You are talking about dropped assembly, 4 talking about misloading in the core, or both?
5 HR. CARROLL: Or loss of cooling of an assembly 6 during, during refueling operation.
7 HR. LANDRY: The refueling plan for this reactor 1
8 would be for the assembly, and I wi?.1 get into this more in !
9 the discussion of Chapter 4. For the assembly to stay in core 10 for normal time would be 20 months. The total--excuse me.
i 4 11 The total life of the fuel assembly would be 60 months.
I 12 Refueling would occur every 20 months. For one fuel cycle, an a
I 13 assembly would stay within the reactor vessel, and would 14 receive sodium coolant while the fission products decayed, 15 while the assembly cooled after that period of time, the 16 assembly would be dry handleable, so it would not be necessary
- 17 for it to stay within the cooling. .
i 18 HR. MICHELSON: Question on external events--you say l 19 it is consistent with those imposed for light water reactors.
20 Since light water reactors do not have liquid odium present, 21 how are you going to view the potential hazard, fire hazard l 22 with the sodium?
23 HR. LANDRY: We will consider that in addition to i
l 24 the external events.
() 25 HR. MICHELSON: You will have to postulate some kind HERITAGE REPORTING CORPORATION -- (202)628-4888
102 1 of maximum credible release of sodium and credible release of 2 water in conjunction with it and so on, or just air burning.
O 3 You are going to have to do quite a bit of thinking before you 4 decide how big a fire you might have. ,
5 HR. KING: Sodium fire we are not calling an 6 external event. That is part of internal events that we 7 already looked at.
8 HR. MICHELSON: The Agency givas me a problem of, ,
i 9 problems because each part of them defines external event in 10 their own way. You say no, that is not an internal event I 11 guess, and other parts of the same Agency say no, fires are 12 always external events, so get your act together a little bit
- 13 so I know how to term it. Whatever you want to call it is i
() 14 fine.
15 HR. KING: We have already looked into sodium fire 16 aspects of design. What we are talking about on external 17 events in a vugraph is things like earthquake, floods, that 18 sort of thing where we are going beyond the design basis for
< 19 LWRs. and whatever they come up with for--
l l 20 HR. MICHELSON: I can ask my question directly i
{ 21 instead of saying when you get to it. Okay.
22 How did you handle what is the maximum credible i 23 release of sodium for fire purposes?
24 HR. KING: Well, for the primary system, it is the l
25 critical leak in the reactor vessel. For the--
[}
HERITAGE REPORTING CORPORATION -- (202)628-4888
103 ;
1 HR. MICHELSON: That's an unisolated leak?
2 HR. KING: That's an unisolated leak. For the 3 intermediate system, it is the moderate energy flood system 4 break in a pipe that duicps sodium into the, into the runway 5 where the pipes are, the intermediate system pipes are, or 6 into the steam generator building, the steam generator.
7 HR. MICHELSON: Moderate energy break also?
8 HR. KING: Hoderate energy system break size.
9 HR. HICHELSON: And when, what provisions for 10 protection against that fire do you envision?
11 HR. KING: The pipe-- ,
12 HR. HICHELSON: Or did you postulate there was no 13 mitigation, just have draining and kept burning? That's t
() 14 another possible model. I don't know which one you nad to 15 choose.
! 16 HR. KING: The model assumes you dump. At the rate
- 17 you are dumping the sodium, it all dumps out. The inventory 18 of the intermediate dumps out.
19 HR. MICHELSON: Did you postulate all of that 20 combusted?
21 HR. KING: I don't know how much combusted, but it 22 is steel line cell, and--
23 HR. LANDRY: Steel line cell, it is guard piled 24 piping until it gets into a steam generator compartment.
(} 25 HR. MICHELSON: You do have an acquired heat release HERITAGE REPORTING CORPOR.'. TION -- (202)628-4888
104 1 while it is reacting?
2 HR. LANDRY: They do have fire protection systems.
}
3 They do have catch pans.
4 HR. MICHELSON: So you have already done the fire 5 analysis and concluded that fire is not a problem?
6 HR. KING: Sodium fires are using the same 7 technology Clinch River and FFTF used and demonstrated, and we 8 don't see sodium fires as any more hazardous in this plant 9 than they were at Clinch River.
10 DR. KERR: In those cases, though, wasn't there a 11 limited supply of oxygen inside the containment structure?
12 HR. KING: Only in the cells that had contained 1 13 primary system pipe; those were looped plants. This is a pipe
( 14 plant. The only place for primary sodium to leak out is into 15 the space between the vessel and the guard vessel, and that is 16 an inert space.
17 HR. MICHELSON: There is circulating sodium as well, 18 the heat exchangers?
19 HR. KING: The sodium that is outside the reactor 20 vessel is intermediate sodium. It is not radioactive.
21 HR. MICHELSON: I am not worried about 22 radioactivity. My question is fire, fire effects, which might 23 ult.mately lead to other problems, but that's the fire, the
- 24 one outside.
() 25 HR. KING: In terms of looking at the intermediate HERITAGE REPORTING CORPORATION -- (202)628-4888
105 1 system spilled, yes, that was looked at. Sodium fire
(} 2 protection feature, the leak detection features were 3 essentially the same as Clinch River.
4 MR. MICHELSON: Were there provisions for relieving i
4 5 the build-up of pressure in those, in the, those lined 6 chambers that this piping goes through?
7 MR. KING: Yes.
8 MR. MICHELSON: Where does it lead to?
9 MR. KING: Is it catch pans? -
10 MR. MICHELSON: Relieving pressure to--you can't 7
1 11 relieve pressure to a catch pan. The fire creates, heats the r 12 gas. That expands. It has got to go somewhere. Where does
- 13 it go to?
14 MR. KING: *nbably goes outside. I would have to--
15 MR. LANDRY: One compartment to another compartment; t
16 can eventually find its way outside. They are not sealed. ,
1 q 17 These are not sealed compartments.
18 MR. MICHELSON: But it is assured that the i
19 ventilation is outside and not into areas where you would not 20 be able to tolerate the heat? !
- 21 MR. LANDRY
- The ventilation would not carry it back f 22 into the reactor silo or back into the head access area, or 23 into the equipment vaults.
24 MR. MICHELSON: Thank you.
() 25 CHAIRMAN WARD: Ralph, before you leave this, the HERITAGE REPORTING CORPORATION -- (202)628-4888
106 1 external events, those imposed on LWRs, and you said to be 2 imposed on LWRs, I guess I'm, I don't know what you are O 3 talking about there.
1 4 HR. LANDRY: Those that are to be imposed on LWRs, 5 we are aware that for the severe accident implementation plan 6 for the future LWRs, that the external events have not been 7 defined, and there is still research going on, on the possible 8 external events, and we are saying that we are holding off on 9 defining the external events which must be considered for the 10 LHR pending that informa'cion, so that what we impose will be 11 consistent with those impcsed on the LWRs.
12 CHAIRMAN WARD: Okay. I'm not sure I understand a
13 that. Do you understand that?
() 14 HR. WYLIE: It has to do with the defining external 15 events if we conducted the PRA?
16 HR. KING: There is an IPE uorking group. It is a 17 working group established to define what severe external 18 events do you consider in the IPE, build upon that.
19 CHAIRMAN WARD: Okay.
20 HR. CARROLL: Fifteen months or something?
21 CHAIRMAN WARD: That's what you are talking about 22 for the IPE?
23 HR. LANDRY: That's right. I meant what I said--the i
24 severe accident implementation plan.
i 25 CHAIRMAN WARD: Okay.
i HERITAGE REPORTING CORPORATION -- (202)628-4888
107 1 MR. MICHELSON: I would like to ask a question on 2 the response that we have just heard. The IPE program for the 3 external events aspects of the IPE programs will include fire?
4 MR. KING: Yes. That is one--
5 MR. MICHELSON: As opposed to your case, you are not 6 going to include it? You have already included it. Let me 7 get to what my concern is.
8 In the process of deciding how to handle fire 9 related affairs, they will come up with some kind of criteria, 10 whatever for external events under the water reactor program.
11 Are you going to go back and look at those, whatever 12 comes out of that, so it will cesult and go back to see if you 13 have handled it for the--
() 14 MR. LANDRY: We will look at what comes out and 15 determine what is applicable and what we have encompassed or 16 enveloped already.
17 MR. MICHELSON: If there is anything new that you 18 haven't considered that might be for you to go back and 19 backfit it?
l 20 MR. LANDRY: That is why we say consistent with, why 21 we don't say we will impose directly.
22 MR. MICHELSON: But you kind of assured me you have 23 already done it and it is hard to be consistent with it when 24 you don't know what they are going to come up with.
25 MR. LANDRY: We will review what they come out with.
HEFiTAGE REPORTING CORPORATION -- (202)628-4888 i
108 1 MR. MICHELSON: Thank you.
(} 2 (Slide) 3 HR. LANDRY: I would like to move on to the core 4 design. As you heard earlier, the plan for the PRISM reactor 5 is to use a metal alloy core rather than oxide. The alloy 6 which we have reviewed so far has been a tertiary alloy 7 consisting of uranium, 26 percent plutonium, 10 percent 8 zirconium. These are for the fuel assemblies; the blanket 9 assemblies for uranium, ten zire alloy; all assemblies, all 10 ducting, all materials internal to the reactor, are clad with 11 HT-9 steel.
12 This briefly lists the core's configuration. I 13 don't want to go through all the details of that, simply to
( 14 show that the plan--of course, you have to look at this. This 15 lists both the metal core and oxide. We have not reviewed an 16 oxide core, so you just disgard everything down here.
17 CHAIRMAN WARD: Why don't you lower the slide?
18 Thank you.
19 HR. LANDRY: The core that we have looked at uses an 20 annular design with the fuel assemblies in a double annulus, 21 with six control rod assemblies. The design would have within 22 each of those fuel assemblies approximately 331 fuel pins.
23 Reviewing the core design--
24 (Slide)
() 25 HR. LANDRY: The criteria which we have used are HERITAGE REPORTING CORPORATION -- (202)628-4888
109 1 that the fuel is not damaged by either normal or abnormal 2 oporation. That includes anticipated transients. We reviewed 3 the fuel design with the anticipation that there would be no 4 fuel melting, no sodium voiding, fuel damage would not prevent 5 control rod insertion even if there was fuel damage; number of 6 failures that have been estimated to occur such as one beyond 7 clad reach are not underestimated, and that there is always 8 coolability maintained in the core.
9 (Slide) 10 HR. LANDRY: Just for information purposes, the 11 plant that was submitted says that it may incorporate on-site i 12 fuel reprocessing that would incorporate fabrication, ,
13 reprocessing, waste management. That you heard Mr. Griffith
() 14 explain this morning. We did not review any on-plant 15 reprocessing plants at this point.
16 We have looked at the pyrometallurgical technique
! 17 that has been developed at EBR 2. We do note that that ,
r 18 technique has been demonstrated in a laboratory scale. It has 19 not been done atraight on the commercial scale to date.
- 20 Composition of reprocessing fuel has not been defined at this r i
21 point. These are areas which we are simply highlighting as l 22 having been mentioned in the design. They have not been !
23 presented to us at this point. We p an to review this in the 24 near future, or begin a review of it in the future.
() 25 DR. RE11CK: Realizing you did not consider HERITAGE REPORTING CORPORATION -- (202)628-4888
110 1 reprocessing, does the staff feel they know what Part 50 2 requires for reprocessing?
O 3 Any time I have read Part 50, I get awfully confused 4 on what applias to reprocessing, what applies to reactors. I 5 thought it was, long ago it would be important to try to 6 separate out, address reprocessing questions separately kind 7 of like Part 72, pull out Part 50 for the understandability of 8 everybody.
9 MR. LANDRY: It was one of the things we have 10 discussed with the Office of the General Counsel, with the 11 lawyers, that if we start reviewing the reprocessing plan, if 12 there is a formal proposal to incorporate reprocessing, we 13 would have to go back and review our regulations and their
(} 14 applicability for reprocessing and permitting reprocessing in 15 particular on site with the reactor.
16 DR. REMICK: Has anybody identified that now might 17 be the time to start that?
18 MR. LANDRY: We are planning to start that in FY 19 '89. That's part of our program. At this stage, we have 20 reviewed the first PSID for PRISH and for SAFR, You have 21 heard this morning that DOE has made the decision that General 22 Electric will continue with the next round design. T?.e plan 23 is in approximately April I believe it is of 1989, or 1990, 24 the next round of PSID would come in. In the interim period, 25 we plan to look at the reprocessing plans that have been O
HERITAGE REPORTING CORPORATION -- (202)628-4888
111 1 developed to date and start getting an idea, an understanding i i
2 of what has been designed.
3 DR. KERR: You do not say anything about possible 4 recriticality in the upper lumen because you assume fuel 5 damage that would be sufficient to produce that as incredible?
6 HR. LANDRY: That is correct.
7 (Slide) 8 HR. LANDRY: The cladding term which has been 9 proposed for the PRISH core utilizes HT-9 steel as the 10 cladding. Now we notice that from the information avsilable, 11 HT-9 has good swelling characteristics; with the irradiation, 12 does not swell.
13 However, this material is relatively new and there
() 14 is very little data available. The Subcommittee was at EBR 2 15 last year in August and had a presentation on this entire 16 process, the cladding, the metal fuel, the pyrome*allurgical 17 process, should be aware of some of these concerns.
18 We have been involved in reviewing the research 19 program, the R&D program for the metal fuel, and we have noted 20 that there are little data available for, on the 21 thermal-mechanical properties of the material, that material 22 properties which we have seen were developed from one single 23 batch of cladding material. On that basis, we feel that we 24 need to see a batch-to-batch consistency. W9 need to see 25 material properties for more than one single batch.
HERITAGE REPORTING CORPORATION -- (202)628-4888
l 112 1 The welding fabrication techniques for this material 2 are relatively new. We have to understand more about the I
O 3 techniques they are planning.
4 Our concern which we highlight here is that there is 5 a need for a data base on this cladding material. However.
6 while we say that, we also point out that there is a large R&D 7 effort underway sponsored by the Department of Energy and AEOD 8 reviewing that R&D effort, the plans for the next three to 9 five years. We feel that the data base can be provided if 10 that R&D effort proceeds as currently listed.
11 We have a great deal of confidence that it will be l
12 possible to answer the concerns and provide information which 13 we will identify, so the information which we have seen to
(} 14 date regarding the fuel performance, the life limit on the
- 15 fuel is based on creep ruptures of the HT-9 cladding from j 16 pressurization of the fission gas plenum.
I j 17 We have seen, however, that there are favorable i
! 18 conditions such as the fuel, the metal fuel is highly porous.
l 19 Now of course, we will talk about down here that that does
( 20 cause some other problems, but basically, the high porosity of i
21 the fuel is a favorable point. The HT-9 cladding does have l
I l' 22 high strength, no swelling characteristics, and there is very
\
- Jacribed earlier 23 good fuel coolant compatibili l
l 24 this morning by Mr. Griffith 25 The research prog' -
+ v will l
1 1
HERITAGE REPORTIN3 C l f \
1 1 r
113 i study local effects along the pin axis, and in particular such t 2 items as fixed gas retention and swelling of the metal fuel
( That program is also underway to study the species 3 itself. l l 4 migration, and with burnup. This is a concern to us because j
. 5 of the porosity, because of the species migration. We see a j
6 great deal of fuel restructuring underway in the range of one i
- 7 and a half to two atomic percent burnup of fuel. l l
, 8 Specifically what occurs in this range is the i 1
l 9 excellent thermal conductivity which metal fuel exhibits ;
i :
I 10 decreases in the one and a half to 2 percent burnup range. (
) k l 11 However, the experience to date has shown that as burnup '
- 12 increases beyond that point, when it gets beyond 3 percent, 4 i l 13 the conductivity again comes back up. We feel that there is a [
1 J ,
i i
1
() 14 great deal more burnup information necessary to substantiate j 15 that information. [
i l 16 On the plus side, while the restructuring is going j I li on, and the fuel thermal conductivity is decreasing, the 6
1 18 porosity links up, and we are seeing a backflow of the sodium j
, L
) 19 bond material into the porosity, and that may, basad on some J I t
! 20 preliminary information, result in a favorable increase in the ;
I 21 thermal conductivity of the fuel, k 22 A problem which we have identified is with the metal !
! i l 23 fuel, and the HTI cladding, a low fuel clad eutectic j 6
l 24 terperature on the order of 800 degrees. This is a concern in l
{
25 that with such a low clad eutectic tenperature, you can't get !
O !
I f
HERITAGE REPORTING CORPORATION -- (202)628-4888 l i
114 1 into damage of the fuel assembly or fuel pir at a relatively d
(} 2 low temperature.
3 On the plus side--
i
! 4 CHAIRMAN WARD: That is 800 deg' tees with what-- p i ,
5 HR. LANDRY: I believe that is 800 C. I should have j l
6 said C, not F. ,
a 7 on the plus side, you realize that with oxide fuel, !
i l 8 when you go into fuel melt conditions, you are at extremely l 4 l 9 high temperatures, getting up to temperatures above 3,000 t 10 degrees Fahrenheit. That gives you a very, very high (
t 11 temperature material to try to control and to see moving {
12 around inside the reactor vessel. [
! 13 If we get into a fuel melt situation here, it is at
- ( 14 low temperatures, and low temperatv.re material you see moving i i 1 15 around the reactor vessel, so we see a plus and a minus. On ;
I i j 16 the low fuel clad eutectic tempers.ture you get into trouble l 17 sooner, but the trouble you get into may be even more easily 18 controlled.
(
19 (Slide) !
l h j 20 HR. LANDRY: Reactivity control system, which has l
f 21 been proposed for PRISH, utilizes six active control rods.
i 22 Those six rods all have independent control systems, }
f
, 23 independent electronics, indepondent power channels, j f
- 24 independent motor drives. They are all safety grade. The 1
l
() 25 latch system which was developed or which has been proposed is ;
f 1 :
l l l HERITAGE REPORTING CORPORATION -- (202)628-4888 j
115 ,
j 1 that which was developed on the Clinch River breeder reactor, ,
i ;
I 2 has had a great deal of development and R&D offort on it.
O 3 The control rods are equipped with a second motor.
4 s 4 They have a small motor which results in driving the rod out, 5 and driving the rod in under normal conditions. They also 1
! 6 have a large motor which is capable of exerting 2,000 pounds i
i 7 of driving t'crea on the control rod. Should a trip occur, the 8 rod would trip and the drive line would be driven in behind l 9 the rod with 2,000 pounds of force. This would as proposed j 10 tend to force the rod into the core irrespective of any 11 relocation of fuel.
12 That motor in safety grade. The control system for i
13 it is on the 1E bus, and the motor drives in one direction 14 only. It is designed to only drive in. It cannot withdraw l 15 the rods. It is considerably higher power motor than the l
) 16 motor which pulls the rods, so should the rod pull, with the I
17 motor trying to draw the rods out after a trip, the driving l
i 18 ;totor would counteract that and force the rods in in spite of l
j 19 withdrawal.
1 i 20 tiow earlier when we talked about the bounding 21 events, we did not give credit for the drive in.
l 22 CHAIRHAll WARD: When you say something like that, it i
) 23 would drive it in irrespective of any relocation of fuel?
24 HR. LAllDRY: Which we have seen to date.
1
! 25 CHAIRMAN WARD: You know, there is probably a broad HERITAGE REPORTING CORPORATION -- (2021628-4888
116 1 spectrum of possibilities of what the fuel in going to look 2 like if it re?.ocates, and I mean I will grant you that this 3 might very well drive it in for much of that spectrum, but to l i ;
i 4 say, you know, such an absolute statement seems questionable. l
! l 5 HR. LANDRY: I qualified it by saying what we have :
I
! 6 seen to date.
7 DR. SIESS: There must be some configuration even if l
i j 8 you go within, it didn't shut it down? ;
a 9 HR. LANDRY: Well, the control system, or the ,
I i
) 10 control rods for the reactor have sufficient negative i 1 11 reactivity to shut the reactor down with only one rod, 1 12 DR. SIESS: No matter what the geometry of the fuel
! I i 13 is? f O 14 "R. tANDRv: I can e saY no matter what the i f
1 15 ocometry. Of course, if the rod cannot get into the core ;
16 material, then it doesn't matter how much it is worth. In f
l-17 other Mords, if--what you mean by no matter what the geometry, [
18 if the core :taterial was completely separated f rom the control [
i 19 rods, such that the control rod would miss all of the core !
20 material, of course it would not have any effect. We haven't (
21 looked at that.
23 HR AVERY: What was the maximum positive reactivity !
23 that you can get on rod withdrawal? f l
24 HR. LANDRY: Thirty-five cents, f l
25 HR. CARR0LL: Is it designed so only one control rod HERITAGE REPORTING CORPORATION -- (202)628-4888 [
_ . - . - - - . - - -a
l l
117 1 can be withdrawn at a time?
2 HR. LANDRY: That's right. The control system is c
3 designed so that one control rod is stepped out part way. ;
4 Then another control rod can be stepped out part way, and )
5 working, the operator can work around all six rods, but one 6 rod cannot be withdrawn--they may want to correct me--I ,
7 believe one rod cannot be withdrawn completely at one time, !
8 and no more than one rod can be in motion at one time. ,
9 DR. GYOREY: There is only a single power supply I l
10 which is switched to any of the six rods, so you can only I 11 apply power to a single rod at a time. ;
12 HR. LANDRY: The concern that we have with the 13 system is that there is only one single active shutdown
(} 14 system. As you will recall, the GDC specify two active 15 shutdown systems.
16 DR. SIESS: What is your concern, single or active i
17 or both?
18 HR. LANDRY: Single active. !
L 19 DR. SIESS: Suppose there were two active systems? ;
20 That would be all right? Or single passive system, would it ,
I 21 be all right? You use two adjectives. I am wondering which ;
t 22 one concerns you. }
23 HR. LANDRY: That's just the point that we are j t
24 trying to nake. We are concerned, we are looking at this j I
25 ~
m beaause there is one single active shutdown system. i O
T!lAGE REPORTING CORPORATION -- (202)628-4888 I
118 1 There is inherent shutdown capability of this core. Now we 2 have looked at this. We are giving, at this point, we are
{
3 saying that the design is--
l 1 4 DR. SIESS: If there are six rods, any one of them l
l 5 will shut it down, isn't that six active systems?
l 6 HR. LANDRY: No. They are not diverse.
7 DR. SIESS: I didn't say--I mean redundancy and 8 diversity are two different things. A single implies nothing 9 about that. Active implies nothing. I am trying to 10 understand what your concern is.
11 HR. LANDRY: I should have made that clear. We mean 12 the single redundant active, single system which is neither 13 redundant or which is not diverse. It is redundant. There
() 14 are single systems.
15 DR. SIESS: It isn't what you have. It is what you 16 haven't got that bothers you?
17 HR. LANDRY: Right.
18 DR. SIESS: You don't have another--
19 HR. LANDRY: We don't have another active system.
20 DR. SIESS: Don't have another system. You would 21 accept another active system if it were diverse? You would 22 accept another passive system which would have to be diverse 23 if this was active, right?
24 HR. LANDRY: Right.
() 25 DR. SIESS: Okay. You have got a lot in those three HERITAGE REPORTING CORPORATION -- (202)628-4888
. . . _ _ . . _ j
-. -- .. . - . - _ ~ . - -
119 1 words there.
{} 2 HR. LANDRY: I was trying to make~ bullet items and 3 make this quick.
4 On the favorable side, however, when we, after we 5 make that statement, we do look at the system that has been 6 designed. We sa/ that one, the control rods are safety grade.
7 Control system 3s safety grade, and two, there is a safety 8 grade drive-in function for those tods. Not only is there a
9 safety grade trip, but safety grade drive-in function. Then 10 we add on to that the inherent characteristics of the cover. ,
11 We, we have to figure out how we get around the requirements 12 of the GDC.
13 DR. KERR: The characteristico of the core don't ;
() 14 shut it down, do they?
! 15 MR. LANDRY: They will take it to a subcritical 16 condition.
17 DR. KERR: How does, how does it ever become 18 critical then with all control rods out?
19 HR. KING: Takes it to lov power critical condition, 1
i 20 elevated temperature, low power.
21 DR. KERR: How do you ever get power out cf the 22 reactor?
i l 23 MR. KING: This is beyond normal operating 1
24 temperature. You would have say a loss of flow. The core
() 25 heats up. You get thermal expansions, t?.e grid plate--
l l
HERITAGE REPORTING CORPORATION -- (202)628-4888 i
l ~ _ - , . . _ . _ , . , . _ , . - _ . . . - _ _ , _
._..,-,,_,_...__._,.m,,,..,_.m _ .
120 1 DR. KERR: You are saying--all right.
2 HR. KING: Tends to shut the power down, but the 3 plant sits there critical.
4 DR. KERR: You are assuming that no heat is being 5 removed then?
r 6 HR. KING: Heat is being removed from that passive--
7 DR. KERR: You are not taking up full load or 8 anything. You have shut down your heat removal, normal heat
! 9 removal system, permitted the reactor to heat up?
10 HR. KI! 'i: Yes.
11 (Slide) 12 HR. LANDRY: Okay. The inherent reactivity i 13 coefficients designed into the core are the Doppler O 14 coefficieat. watch for the meta 1 fue1 is sma11er taaa for 15 oxide, smaller negative factor than for oxide; the positiva
! 16 density coefficient, which results in a rather large positive 17 sodium coefficient for voiding.
l l
, 18 DR. KERR: In the case of the Doppler, is the--what i
19 do you mean by a negative factor? You aren't referring to the 20 temperature coefficient, or are you?
21 HR. LANDRY: Yes. It is less negat've than it would l 22 be for oxide fuel.
l l 23 DR. KERR: Actual coefficient, not just the fact 24 that the temperature goes up a lot more for oxide?
25 HR. LANDRY: Right. There is a negative coefficient l
l l HERITAGE REPORTING CORPORATION -- (202)628-4888
121 1 for axial expansion, and this is a function of burnup. There
. is a negative coefficient for radial expansion, which is used 4,.;
3 or comes into play for the long, slow heatup transients. This 4 is where they, the core expands, the grid plate expands.
5 There is a very slow acting coefficient.
6 DR. REMICK: Qualitatively, why it is due to 7 increased leakage I take it, radial expansion results?
8 MR. LANDRY: Right. So the radial and axial 9 exphnaion coefficients are negative due to increased leakage.
10 DR. REMICK: Leakage due to what?
11 HR. LANDRY: Hoving the fuel aptrt.
12 DR. REMICK: Is it the decreased density inside?
13 Something seems inconsistent there.
j
() 14 DR. KERR: It is increased surface t.o volume rmtio.
15 DR. REMICK: The volume is going up greater than the 16 surface? Am I right? i 17 HR. LANDRY: No. ,
- 18 DR. REMICK
- The surface is going up as basically 19 the diameter, and the volume is going up as the square?
20 MR. KING: Decreased densities.
21 DR. REMICK: Okay.
r 22 MR. KING: Same amount of fuel in a larger space.
i DR. REMICK:
23 Okay.
! 24 MR. LANDRY: Because of the long-- 3 25 DR. REMICK: By the way, I think the SER says l
HERITAGE REPORTING CORPORATION -- (202)628-4888
122 l
' l 1 increased surface. That's what got me thinking about it
! 2 because I think the volume is going up faster than the 3 surface, so it has to be density, but your SER says it 19 the i 4 surfaco.
)
5 HR. KING: I'll check it. )
] I 6 HR. LANDRY: Okay. The long fuel assemblies have ;
l .
7 been designed to prevent unlimited, or limited three bow l 8 condition has been imposed which will keep a bowing
- 9 coefficient negative, force the core to bow outward rather i
10 than inward.
1 11 control rod drive line expansion causes insertion of i
12 the control rods for the long heatup transients, but as the 13 time draws out and you get to a very long heatup, the reactor
() 14 vessel itself starts to expand which then starts to withdraw 15 the control rods. The overall coefficient for the transients 16 which we have been looking at comes out negative when all 17 these are added up.
18 tiR . AVERY: What is the size of the positive sodium 19 void coefficient?
20 HR. LANDRY: Later when Greg Van Tule from 21 Brookhaven gets up and discusses the accident analysis, Craig 22 will have seme charts which show the coefficients and show the 23 size of the different reactivity coefficients.
24 HR. KING: Just ballpark number would be like four
() 25 or five dollars positive reactivity if you voided the whole HERITAGE REPORTING CORPORATION -- (202)628-4888
123 i
i core. Ralph is not talking about sodium densities up there. ;
L
[ 2 HR. AVERY: I realize that.
j 3 DR. SIESS: By voiding, you me&n the sodium is j 4 actually not there? It is replaced either by gas or
{
5 something?
4 6 HR. LANDRY: Right, sodium vapor.
j !
7 DR. SIESS: Sodium vapor. !
I f
8 DR. KERR: Now the fact that total coefficient. I I 1 9 guess you mean with temperature is negative, is not l t
9
- 10 necessarily of sufficient comfort because you can certainly f i i 11 get severe power oscillations, i i i
12 HR. LANDRY: Right. ;
i 4 13 DR. KERR: Even under those circumstances; you have 14 looked at that presumably from what I have read, but in order 15 to look at it, you have to know what the sodium void is. ;
i 16 HR. LANDRY: It does give us a concern at this
{
f 17 point, and we are concerned that there is a positive sodium f 18 voiding or sodium void coefficient with this core, even though j i
19 the transients we have looked at so far, we have seen an I i
l 20 overall coefficient that stays negative. We are not '
t 21 completely comfortable with a void coefficient which is [
i 22 positive. !
I 23 DR. KERR: You will recall DBR 1 one got an overall f
24 coefficient of negative. !
l 25 MR. AVERY: I am a little confused by that last HERITAGE REPORTING CORPORATION -- (202)628-4888
124 1 statement if it is really correct. I think what you are 2 saying is the overall coefficient is negative as long as the i O 3 only sodium effect is change of density with temperature, but ,
4 if you actually void, then I assume the void would override 5 everything else, f 6 HR. LANDRY: Then it becomes strongly positive.
7 HR. KING: You are going to hear this afternoon from .
i 8 Brookhaven some analysis we have done that looks at the ;
9 reactivity effects, the varicus components. Haybe we can come 10 back to it then. We can put some numbers to Some of these {
11 words, some of this oscillation concern.
12 CHAIRHAN WARD: Will that, for example, give us a 13 picture of how important the Boeing coefficient is? I r
() 14 HR. LANDRY: Right. Well--
15 CHAIRHAN WARD: And what in the design assures i t
16 outward bowing instead of--I will wait for that.
l 17 HR. LANDRY: Craig will give some information on 18 that.
19 CHAIRHAN WARD: Good.
20 DR. REMICK: When they address that this afternoon.
l r
21 alert them if they could give us some hint qualitatively of [
22 what a designer might do to make that positive coefficient 23 less positive, or neutral.
24 HR. LANDRY: We have been discussing that with the 25 designer and with Argonne. We hava been discussing what i
HERITAGE REPORTING CORPORATION -- (202)628-4888 5 I
125 !
1 methods can be used to reduce the positive coetficient, and GE !
1 l
{} 2 has heard us and they are concerned about it, and in the next 3 round they have already mentioned that they are looking at f l
4 means by which they can reduce that positive coefficient.
5 DR. REMICK: Okay.
l 6 HR. LANDRY: With regard to the reactivity l 7 coefficients, the independent analyses which we have seen have l
4 8 shown a consistency with those from the designers. We are l
l 9 concerned with relatively high uncertainties in the reactivity
- 10 coefficients. We are aware that there is more information 1
11 coming out of FFTF which may help reduce that uncertainty l
12 level, but we are waiting to see that information, he are i 13 concerned that a very fast flow reduction could lead to sodium
() 14 voiding.
I 15 A favorable factor in all of this is again the i~
16 prototype test program. The prototype test program is l
17 expected to perform a number of these transients and obtain
) 18 the data which would substantiate these reactivity i
l 19 coefficients.
I l 20 CHAIRMAN WARD: Including data on voiding?
21 HR. LANDRY: We hope not, but if it occurs, we would i
22 not turn the data away, l
f 23 HR. KING: There is a critical experiment assembly i
j 24 plant which we have done, could loor at voiding. I don't know r
f () 25 the details of what all the measurements are going to be, that HERITAGE REPORTING CORPORATION -- (202)628-4888
126 1 one, one way of looking at voiding--
{} 2 DR. KERR: ZPPR.
3 MR. KING: ZPPR.
4 MR. LANDRY: And TREAT tests that are planned which 5 will provido us with information, as we discussed aarlier, 6 with regard to the active shutdown system.
7 We do have a question regarding inherency. Can we 8 regard inherency as a shutdown system in itself? And second, 9 can we accept a design that does incorporate a positive 10 reactivity coefficient even though the overall coexficient 11 might be negative?
12 A few of R&D issues, and here let me caution that we 13 don't really mean that these are issues that we are at odds 14 with the designer on these. These are simply areas which wo 15 feel strongly enough about to highlight.
16 The fuel, we feel that the low outectic temperature, 17 the sodium logging, the thermal properties with burnup, the
{
18 cladding properties, the slow overpower behavior--
19 DR. KERR: What is sodium logging?
20 MR. LANDRY: That's the backflow or--yes. For 21 better, for lack of a better term, Lackflow of the sodium into 22 the porosity of the metal fuel as the fuel burns up; the bond l 23 sodium--I didn't show it. I didn't have a diagram.
24 DR. KERR: That's enough.
() 25 MR. LANDRY: Inside the fuel rod, the metal fuel has HERITAGE REPORTING CORPORATION -- (202)G28-4898
127 1 sodium placed in the fuel rod at manufactured, so that during 2 the early stages before the fuel itself expands, or as well as
)
3 to contact the clad, there is good thermal contact made 4 between the metal fuel pin and the cladding by a sodium 5 backfill. That sodium, when the fuel burns up and goes 6 through this period of low conductivity and high porosity, the 7 sodium can backlog or backfill that porosity.
8 The R&D program we feel as it is underway can answer 9 these questions, and we encourage fulfillment of that program.
10 We do have some concerns about fuel failure 11 propagation or r'in beyond clad breach conditione, but again, 12 these are parts of the fuel research program that is now 13 underwak.
() 14 Reactor physics, concern is the ree.ctivity feedback 15 confirmation that will be obtained through the separate test, 16 the ZPPR test and the TREAT test and again through the 17 prototype test program--
18 DR. SIESS: You mentioned the fuel propagation. Are 19 there any provisions for detecting fuel variance?
20 HR. LANDRY: Failures, delayed neutron monitors 21 within the reactor vessel, those monitors would detect fuel 2.' failures, or cladding failures in particular.
23 DR. SIESS: Anywhere?
24 HR. LANDRY: They would be able to--at the present r
() 25 time the present design would not pinpoint the exact pin. It HERITAGE REPORTING CORPORATION -- (202)628-4888
128 1 would give a location of fuel assembly, the different fuel
(} 2 assemblies. As we heard, the proposed design would 3 incorporate tag gases within the, urdque to each assembly or 4 groups of assemblies. Should those tag gases be released, 5 they would be detected by the gas sampling system.
6 DR. SIESS: You have just got a leak, but if you 7 have got a failure, you want to shut down before it leaks, you 8 haven't got time to look for the gas analysis.
9 MR. LANDRY: No. The fast failure I don't believe 10 would be detected, 11 MR. AVERY: The delayed neutron would pick that up.
12 The only problem there would be is that you have to assume an 13 outlet blockage. Then you k'now where that, how is that going
( 14 to get--
15 MR. LANDRY: You would also have to consider transit 16 time to the dete-tor. It would not be an instantaneous 17 indicator of a damage.
18 MR. AVERY: The sodium flow.
19 DR. REMICK: You indicated delayed neutron monitors 20 would begin the reactor vessel or--
21 MR. LANDRY: Yes.
22 DR. REMICK: Chilled them adequately so that you 23 aren't swamped out? Or is it taking off the stream of the 24 cover cas, of the cover gas clean-up system they are looking
() 25 at?
HERITAGE REPORTING CORPORATION -- (202)628-4888
129 1 MR. LANDRY: Let me call on Frank Tippets from GE.
2 MR. TIPPETS: Yes. We have two methods actually.
3 Dr. Landry mentioned one, the neutron monitor, and those are 4 located in the intermediate heat exchanger which are for other 5 purposes, clad, a shield, shield, ed and all the primary fuel, 6 of course, will go by the other method which is pick up any 7 just leakers as fission gas monitor that is located in the 8 tank, and then of course the tag gas.
9 DR. REMICK: Thank you.
10 MR. LANDRY: That takes care of what we had planned 31 on for Chapter 4.
12 CHAIRMAN WARD: Let's see, Ralph. We would like to 13 finish your section by is 12:30; try to help.
() 14 MR. LANDRY: I'll race. Let me turn to the reactor 15 module itself. Reactor module or the reactor vessel itse2f, 16 consists of the silo, the collector plate around the reactor 17 which is a part of the RVACS system, the containment vessel, 18 and a reactor vessel.
19 The entire reactor coolant system, the entire 20 primary coolant system is contained within thu . actor vessel.
21 That means the core and fuel assemblies are contained at the 22 lower part; fuel rod drive, the intermediate heat exchangers, 23 and the reactor coolant pumps are all contained within the 24 reactor vessel.
(} 25 There are no penetrations of the containment vessel HERITAGE REPORTING CORPORATION -- $202)628-4888 e-- -
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130 1 at all. The only penetration that occurs are through the 2 upper head region. The upper head region contains a rotatable 3 plug used for refueling operations. All the systems, the 4 sampling systems, the rotatable plug, all connections, are 5 under normal operation seal welded so that there is no 6 possibility for leakage, or there is a vastly reduced 7 possibility for leakage.
8 Again, the layout for the overall system--
9 (Slide) 10 HR. LANDRY: Is for the reactor coolant system, the il primary system to exchange heat with the intermediate heat 12 transport system, the intermediate heat transport system out 13 to the steam generator, and the steam will drive the turbine.
() 14 DR. REMICK: How are you going to remove the fuel?
15 MR. LANDRY: The fuel would be removed through the 16 rotating plug or rotatable plug in the upper head.
17 DR. REMICK: Into a shield I assume?
18 HR. LANDRY: This would be removed into a shield 19 which would be placed over the upper head region.
20 DR. REMICK: How do they get it out of the reactor 21 sites? Shipped off-site rather than--how do they do that?
22 MR. LANDRY: Shipped in a shipping cask.
23 DR. REMICK: Just a plug up above with no roof over 24 it?
() 25 HR. LANDRY: There is a plug, and then there is the HERITAGE REPORTING CORPORATION -- (202)628-4888
131 1 upper head region.
2 (Slide) 3 MR. LANDRY: There is the upper head regf.on above 4 the reactor vessel. Then at grade, there is another 5 reinforced concrete slab with a plug that is removed for 6 refueling.
7 DR. REMICK: I am still concerned. There is no 8 building there, and we are taking out fuel from this 9 presumably in some kind of a shielded cask, doing this out in 10 the open.
11 MR. LANDRY: This would be shielded cask that would 12 be load down or be placed over this upper region.
13 DR. REMICK: We are going to be doing that out in
() 14 the weather? There is something I am missing here. Does this 15 fuel machine have a housing or something?
16 MR. LANDRY: Yes.
17 DR. REMICK: Okay. The workers are not going to be 18 doing it in January in Michigan in the snow storm?
19 MR. LANDRY: Let's see. Frank Tippets has--
20 DR. KERR: Snow load you won't worry.
21 DR. REMICK: You are right.
22 DR. SIESS: You are right.
23 DR. REMICK: So we do have a building.
24 (Slide) 25 HR. L A!! DRY : The plan would be te have an enclosure HERITAGE REPORTING CORPORATION -- (202)628-4888
132 1 which moves it on the railroad tracks which yoit saw in the 2 earlier conceptual design. The enclosure would contain the 3 refueling machinery which would after opening the refueling or 4 the rotatable plug, move the fuel within the reactor vessel, 5 from the reactor core to its storage location. The fuel 6 within the storage location that has been, that is going to be 7 withdrawn would be drawn up into a fuel transfer cask.
8 DR. REMICK: You can see my crane there. I assume 9 that crane is adequate to handle other things besides fuel, 10 like internals and if for some rcason they will have to 11 replace it?
12 MR. LAllDRY: That's right.
13 DR. REMICK: That is going to be quite an enclosure.
14 DR. SIESS: This whole things moves around to find 15 different . reactor locations, or is there one of these for l
16 each?
l
, 17 MR. LAtlDRY: tio . There are, the design which we l
18 have looked at to date would have one.
19 DR. SIESS: Are those little spur tracks that l
l 20 accommodate this thing?
l 21 MR. LAllDRY: Right. The fuel management plan which 22 we have discussed briefly vith the designer would be for l 23 refueling of one reactor module at a time. The plan would be j 24 to refuel one reactor module, bring it back up to power, 25 transfer the fuel to the reprocessing plant or fuel handling HERITAGE REPORTIllG CORPORATIOli -- (202)628-4888
133 1 storage area, then move on, refuel the next reactor module, 2 work their way down through the nine modules one at a time.
{-)3 u
3 (Slide) 4 HR. LANDRY: Another conceptual drawing of that 5 refueling system--it is hard to see--would be for this large 6 enclosure to move in over the top of the reactor module cask 7 for transfer of the fuel.
8 HR. CARROLL: Would there be fuel in the reactor 9 vessel if I were going to do major maintenance such as replace 10 an EM pump or whatever?
11 MR. LANDRY: We haven't discussed it at this point, 12 out--
13 MR. CARROLL: That would require taking the quote,
() 14 head off the reactor vessel, right?
15 MR. LANDRY: I'm not sure in the reactor if the 16 design at this point would permit removal of such things as an 17 EM pump or not. Let me call on Frank Tippets.
18 MR. TIPPETS: Let me put a, give you another chart 19 here, Ralph, that--
l 20 CHAIRMAN WARD: If you would like to explain it, Mr.
I l 21 Tippets, why don't you go ahead?
22 MR. TIPPETS: Sure. Maybe I can speak loud enough.
l 23 For component replacement, we would use basically the same l 24 scheme as Dr. Landry showed you for the refueling. We would
(} 25 bring the same enclosure over. In fact, this showed the same HERITAGE REPORTING CORPORATION -- (202)628-4888 l . - _ _ _
l 134 1 i
i picture for fuel transfer. Here is for removal of the 1
s 2 electromagnetic pump. We will have a special cask that is i 3 similar to the fuel transfer cask in general outline. It will l I
4 be brought over with an adaptor. That then is a placed over 5 the port and the electromagnetic pump is supported from on the 6 top of the head. The reactor, there is no head that is taken 7 off of the reactor. It is basically this adaptor is put down 8 over the electromagnetic pump, and then the bolts are taken--
9 CHAIRMAN WARD: The top plug can rotate so the hole 10 is over the pump? Is that the idea?
11 MR. TIPPETS: Oh, there is a slab here and that 12 would be removed. There is a slab, in the concrete slab there 13 is a large port. That's the same port that is open when we
() 14 put in the reactor vessel in the first place, We would take 15 that off.
16 CHAIPMAN WARD: Okay.
17 MR. TIPPETS: That's our present plan. ic is 18 certainly, certainly practical to put ports in that concrete t
19 over each component, but at present we haven't done that 20 because it complicates the structure.
1 i
- 21 Our intention is that these are relatively, our 22 design objective is that these are relatively infrequent 23 events, so that we would prefer, at the present time we would 24 prefer to have ,t a littlo bit more cumbersome by taking that 25 concrete port off and then that will give access to the head HERITAGE REPORTING CORPORATION -- (202)628-488b
f 135 1 access area, but at all times the reactor itself will be i
2 sealed. Yes. l 7-)
(_/
3 DR. REMICK: What if you had to replace the IHX?
4 MR. TIPPETS: I think that is more difficult because 5 as you probably know from your prior studies of this design, 6 it is a curved shaped or sectoral shaped device, but the 7 principle would be the same. We would have a special cask for l
8 that purpose that will go down over it.
9 The principal issue that we haven't decided on yet 10 is whether that cask would be cylindrical or the same shape as 11 the intermediate heat exchangers, and future studies will 12 determine that, but the principle is the same as I described 13 for the pump. It wnuld be placed down over that portion of
() 14 the reactor head. The seals would be broken on the IHX and it 15 would be withdraun up into it and removed and then replaced.
16 DR. REMICK: It would be a port through the reactor 17 head sufficient for the IHX to be extracted?
18 MR. TIPPETS: The IHT goes--it is once again that 19 sectoral shaped, and there is a hole in the reactor head of 20 exactly that shape, and the IHX slips down through that, when 21 it is installed, slips doun through that hole. There is a 22 plate at the top that then bolts into place and it is scaled, 23 and that's all part of the IHX uupport structure. It com9s 24 out at as a unit.
25 DR. REMICK: Thank you.
HERITAGE REPORTING CORPORATION -- (202)628-4888
136 1 MR. TIPPETS: You are welcome.
2 MR. CARROLL: Is it fair then to say that the
{
3 reactor vessel would never be opened to atmosphere when therc 4 is irradiated fuel in it?
5 MR. TIPPETS: Absolutely; it will never be opened to 6 the atmosphere. It will always have, always have a, always 7 have inert gas cover on it, and that will be sealed and sealed 8 to--these transfer casks always will be filled with helium.
9 DR. REMICK: Okay.
10 (Slide) 11 MR. LANDRY: Now I realize you haven't had time or 12 haven't had a copy of the material review to look at. The 13 reviewer for that portion was put on other work, and has not
() 14 yet supplied the material, but a couple of the points which he 15 has verbally indicated, concerns he has with the material, 16 that's planned for the reactor module, he is concerned with 17 the use of large bellows on 30 inch pipes. His gut reaction 18 is he just doesn't like large bellows and doesn't feel 19 confident that they can be sealed and maintained properly.
20 He is not particularly pleased with the gimballed 21 joints that are planned. He doesn't know of another mechanism 22 to allow for the thermal expansion, but he doesn't care for 23 the gimballed joints.
24 He has a great deal of concern and dislike for
() 25 ruptured discs, especially very large diameter ruptured discs, HERITAGE REPORTING CORPORATION -- (202)628-4688
_ . ~ . _ _ _ _ _ - - _ . _ _ _ _ _ _ _ , _ _ _ _ _ . _ _ _ _ _ _ _ _ ,
137 1 and he has a concern about, as we had discussed in the
{} 2 cladding material, the lack of data on HT-9 steel. He has 3 pointed out that he has found areas where this material is 4 rianned to bs used over a very wide thickness range. There 5 are locations where there are some pieces that are 22 mils 6 thick, and there are pieces ap to 4 inches thick. His concern 7 is that most materials of this nature are most applicable in 8 particular thickness ranges and he feels that is a rather wide 9 range and he has to be shown that the HT-9 material will be 10 favorable, and can be used over such a wide range of 11 thicknesses.
12 MR. MICHELSON: Where are the large bellows located?
13 MR. LANDRY: They are located on the intermediate
( 14 heat exchange or intermediate heat transfer system piping.
15 MR. MICHELSON: What model of possible failure have 16 you assumed for large bellows? In other words, how big a 17 break?
18 MR. LANDRY: As I say, I haven't received the 19 information from the reviewer, so I don't know what he is 20 looking at.
21 MR. MICHELSON: You mentioned earlier your model for 22 piping was the same as we used in the water case, but the 23 water case never addressed what you do in the case of bellows.
24 The break model, the break mechanisms, everything is quite
() 25 different. Catastrophic failure of bellows is not an HERITAGE REPORTING CORPORATION -- (202)628-4888
138 1 incredible event.
2 MR. LANDRY: I believe that's one of Dr. Johnson's p/
\_
g 3 problems, that he doesn't feel confident in the use of 4 bellows, 5 MR. MICHELSON: I think you have to in the case of 6 bellows assume a circumferential failure of the bellows, and 7 the dumping of the whatever that that causes.
8 MR. LANDRY: The dumping that that would cause is 9 the bellows are placed around intermediate heat transfer 10 system piping. Then the ruptures of the bellows would not 11 result in rupture larger than we have already assumed.
12 MR. MICHELSON: Just guard pipe bellows?
13 MR. LANDRY: Right. Would not result in a larger I
() 14 rupture of the piping system than we have already seen. 7 15 MR. MICHELSON: The piping system itself required 16 bellows also? Or apparently you have an expansion problem 17 with the guard pipe? Don't you have an expansion problem with 18 the pipe itself?
21 pipe is guarding something, presu;mably a sodium pipe.
22 HR. LANDRY: The intermediate heat transfer system L 23 pipe.
l
! 24 HR. MICHELSON: And doesn't it have an expansion 25 problem requiring bellows also?
(
i HERITAGE REPORTIlG CORPORATION -- (202)628-4888
m 139 1 MR. LAllDRY: That's where the gimballed joint
- 2 occurs. The gimballed joint is in the intermediate heat 3 transfer system pipe, and that is surrounded by the bellows.
4 MR. MICHELSOll: Okay. You have to look at that 5 joint also with some kind--it is a different view but it is 6 not necessarily the same failure mode as would be expected 7 from a steel pipe.
8 (Slide) 9 MR. LAllDRY: The concern which wt have expressed to 10 General Electric and which they have said they will, can take 11 into consideration in the next design phase is the isolation
$ 12 of the synchronous machines. As we discussed earlier, the 13 synchronous machines are essential in driving down the EH O 14 pumps shou 1d power se 1est.
15 In the earliest design, the synchronous machines
- 16 were not seismic iso 1ated and we did not feel they were 17 properly separated. That has been changed, and we are waiting 18 for the next round of the design to review it to be sure that
, 19 they are seismically isolated and separated properly.
! 20 (Slide) 21 HR. LANDRY: A couple of specific R&D questions that j 22 we have raised--with regard to the RVACS system, and its 33 operation, the reactor vessel to containment vessel gap is l
l t 24 normally filled with argon. Should a rupture occur and that 25 cap be filled with sodium, there would be a loss of sodium to HERITAGE REPORTIliG CORPCRATIOli -- (202)628-4888
1 the core, even though it would not result i. . w.icovery, 2 would simply be a loss of our reduction in mass in the reactor 3 vessel, but we have seen that some IPEs that credit is taken 4 for improved thermal performance, heat transfer with that gap i i
5 filled. We have a great deal of difficulty analyzing that 6 situation, and feel that further work is necessary, further
, 7 experiments are necessary, further analysis are necessary.
8 CHAIRMAN WARD The thermal performance is improved, i
! 9 but does that mean in the ATWS case that you have to be at a j
j 10 higher power to maintain the temperature coefficients, at high 11 or low power to maintain the temperature coefficients side up?
12 Would the equilibrium power be reached in ATWS be different
]
13 dith this situation?
, () 14 MR. LANDRY: Craig can answer that.
l 15 MR. VAN TULE: Maybe slightly, but I don't think it 1
- 16 is significant.
17 DR. KERR: What happens if that RVACS--I guess its 4 18 channel is filled with water?
l 19 MR. LANDRY: That's one of the conditionr we have 20 looked at is a possible blockage. We haven't said what blocks
! 21 the RVACS from the air flow; simply that it is blocked.
! 22 DR. KERR: I mean what is the effect of having it l 23 filled with water? Does it perform worse, better? About the i
24 same?
() 25 MR. LANDRY: If the RVACS channel is filled with HERITAGE REPORTING CORPORATION -- (202)628-4888
141 1 water, there would be no air flow.
2 DR. KERR: But there would be a steam flow?
)
3 HR. LANDRY: Boil off the steam, boil the steam 4 until you boil the water down far enough to open up that l 5 passage. How heat transfer, boiling heat transfer is a much I 6 more efficient mechanism of heat transfer than radiation. !
7 DR. KERR: Just seems to me that's sort of an 8 obvious way of the thing getting blocked. Somebody is going 9 to look at that I assume, at some point? l i
10 HR. LANDRY: We have discussed the possibility of a 11 flood or some flood of an RVACS channel.
12 DR. KERR: Well, indeed my experience is you can 13 have leakage in. I guess you have some sort of water sealant
() 14 that normally keeps water from coming through the concrete? l 15 MR. LANDRY: We have been looking at the--let me put t
16 this one up.
I 17 (Slide) i 18 MR. LANDRY: We have been discussing with General j 19 Electric the design of the RVACS system which would not only f l
20 keep out water, but which could into an event such as the 21 ontire core was placed down in the silo, it could retain that !
22 sodium without the sodium burning through or damaging the ;
t 23 concreto. l I
24 We have discussed the possibilities of a liner on ,
() 25 this concrete silo such that it can retain the entire core. !
HERITAGE REPORTING CORPORATION -- (202)628-4888 !
. . - _ _ _ _ _ _ a
142 4 1 We have looked at such itens as high lumina content, concrete,
{} 2 which could resist the effects of the sodium which would, 3 could resist the effects of decorium, and we have discussed 4 ways in which that silo we.uld be treated and sealed.
5 (Slide) 6 HR. LANDRY: Let me move quick quickly to Chapter 6.
7 The main concern that we have in Chapter 6 is the
- 8 acceptability of the design without the conventional
- 9 containment. As you heard Tom say earlier this morning, we 10 are looking at the beyond design basis events, the bounding 11 events, and a number of events to try to come to grips with 12 this design without a conventional containment. The design 13 does have a containment, but as you have seen, it is not a 14 conventional containment. This is still an open issue.
15 CHAIRMAN WA!!D: That's an understatement.
16 HR. LANDRY: This is still an open issue with us.
17 It is a concern that we haven't completely satisfied ourselves 18 on at this point.
19 CHAIRHAN WARD: Ralph, I am going to interpret you 20 now, and we will give you 15 minutes more after we get back 21 from lunch, but I would li'a o to take the break now.
22 HR. LANDRY: What do we have to do after lunch?
23 CHAIRMAN WARD: You have got Chapter 7 and 13 24 conclusions, and I would like to give those, you know, a
() 25 decent consideration.
HERITAGE REPORTING CORPORATION -- (202)628-4888
143 1 MR. LANDRY: If you want, I can do Chapter 7 real 2 quickly and we can stop, come back and de Chapter 13. Chapter
}
3 14 I would like to spend a few minutes talking about, 4 prototype test plan.
5 CHAIRMAN WARD: Let's just stop now and come back at 6 1:30.
7 (Whereupon, at 12:30 p.m., the meeting was recessed, 8 to reconvene at 1:30 p.m. the same day.)
9 10 11 12 13
() 14 15 16 17 18 19 20 21 22 23 24
()
HERITAGE REPORTING CORPORATION -- (202)628-4888
1 CERTIFICATE 2
3 This is to certify that the attached proceedings before the 4 United States Nuclear Regulatory Commission in the matter of:
5 Names Advisory Committee on Reactor Safeguards--Subcommittee on Advanced Reactor Designs 7 Docket Number 8 Place: Dethesda, Maryland 9 Date: October 5, 1988 10 were held as herein appears, and that this is the original 11 transcript thereof for the file of the United States Nuclear 12 Regulatory Commission taken stenographically by me and, 13 thereafter reduced to typewriting by me or under the direction 14 of the court reporting company, and that the transcript is a 15 true and accurate record of the foregoing proceedings.
16 /S/ b 4 M t.s 4 l w -
c, d, 17 (Signature typed): Catherine S. Boyd 18 Official Reporter 19 Heritage Reporting Corporation 20 l
21 22 23 24 l
25 1
i
! Heritage Roporting Corporation i (202) 628-4888
, U.S. DEPARTMENT OF ENERGY l NUCLEAR RESEARCH AND DEVELOPMENT PROGRAM 1
! Advanced Reactor Program Presentation to ACRS Advanced Reactor Subcommittee gtNTop
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- 4rEs of Jerry D. Griffith Associate Deputy Assistant Secretary for Reactor Systems Development and Technology Office of Nuclear Energy October 5.1988
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)
I
i o o o NUCLEAR GROWTH SCENARIOS Ehs l y_ _ _ _ -- - __
j 250-5 1
200-q x Upper Reference .: 189 4
l
,e ,..
150- .-
Nuclear ' L ower R6ference i
! Capacity. .
I ' ' .. \
GWe 1
- . ~ .... ' f __,p .$ ' ,,.'116 3
No New - 52 50- Orders O liijiriijiiiijiiiijiiii iiii iiii iiii ;
1987 1990 1995 2000 2005 2010 2015 2020 Source: 1988 EIA Forecast ,, , _
i
, - - - - -_ , _ , _ - _ - - - _ __ g
$ h O O O LMR PROGRAM STRATEGY l
i
i
. t l O O O 1
4
~, EARLY ASSUMPTIONS FOR THE 4
NUCLEAR R&D PROGRAM
, 3.-- --. _ . -.
_ _3 .- - . _
i
)
il NUCLEAFi CAPACITY. GWE Breeders I l
i
- 1500 l 4 3
- tema.wwu (1
.i l 1000 #y ,
1, ,, ,
'4 f
500 7-j' ' '
Demonstration Prog' ram
( #
y Converters j p ,
, o , - , , .
1960 65 70 A 75 80 85 90 95 2000 '
l l }'
' Year of AEC Forecast I
l t
! I
.l l l
I j . -
! O O O l
i l ), REALITIES OF THE 1980s FOR THE l
NUCLEAR R&D PROGRAM xr rr:~_~__2 _ = _ - - - - . _
ar= = _a =_
_=: - m l
l Nuclear ,
i, Capacity. GWE Breeders t
1500 4
Y l
1000 x AEC l Forecast
\ 500 Y Demonstration Program 1 i i T .-
960 65 7 00$
Forecast Ea.ms:
I f
- 1972-77 36 Reactor Ordors CanceDod *1979 TMI Accid.ent l *1973 Oil Embargo *1981 1977 Ban on Reprocessing Roscinded l
- 1977-81 Doublo Digit inflation *1982 Drop in Electricity Domand l
- 1977-83 75 Reactor Ordors Cancouod *1983 CRBR CancoGod
- 1986 Chernobyl Accidont I
4
O O O !
CHALLENGES SHAPING NUCL5AR R&D v=
PROGRAM TO PROVIDE ECONOMIC POWER ==
IN THE FUTURE <
h <
u .
Regulatory l
Public Attitudes
't
- Complexity i
- Plant Life Operations, and Maintenance ilb - Uncertain Load Growth lj 1:
[ .1
- Financing L
li.
t oveazucremo
O O O ADVANCED REACTOR PROGRAM APPROACH l kE l\ p-.---.-.
- x == -
- t 3
CHALLENGES TECHNICAL RESPONSES -
i l Regulatory PASSIVE SAFETY
- t l
l
. Public Attitudes '
MODULARITY
~ ~
i i , Complexity Plant Life, STANDARDIZATION
! ]n Operations and 1 Maintenance IMPROVED WASTE .
R1 MANAGEMENT Uncertain Load I Growth I l Financing
! ^
il ;
4 j ,, 4.. t t 4 +
.1 arrereszuawn j
I
! l 1
O O O
! m=== WHY PASSIVE AND INHERENT SAFETY l
l '
i
- Improves Safety Margin ,
ll
- Requires No Operator Action or External Power
- Leads to Simpler Plants and Simpler Operations j
- Risk of Core Melt Accidents is insignificant i~
j d
- Aids in Plants Capital Investment Protection :
- Aids in Public Perception of Safety h
- l' ll ALL OF THE ABOVE LEAD TO 1 h PLANT COST REDUCTION li >
I p
l l
l
- - - - - - - - - - _ -- , s, _
e o
e O O
'l 1
1 i
i
- LMR PROGRAM DESCRIPTION i
i I
avesnea.ea
.i
- _ . , ,_ -- _w . . - , . __ - __ , - - _
i l
- O O O :
i j --_ _ _ . _ _ . _
2hg ADVANCED LMR PROGRAM APPROACH 1
l
.l y .
)
CHALLENGES TECHNICAL NNSPONEbS PA E S FETY Regulatory _.
- IFR Technology j Public Attitudes MODULARITY Complexity e Factory Fabrication t
- Advanced Instrumentation Plant Life, nd Controls-l
) : Operations and STANDARDlZATION Maintenance
- Design-Certification by NRC Uncertain Load
- Component and System I ',
Growth Refiability IMPROVED WASTE i
Financing MANAGEMENT
- IFR Fuel Cycle
- Actinide Burning i .
l f.
l
O O O g-p,--- METAL FUEL CYCLE POTENTIAL BENEFITS =-
1:
u
- ,
- Improved Fuel Performance
- Greater safety margins ll - Smaller core 0 - Fewer control rods I - Improved breeding potential l!
- Improved Fuel Cycle Economics
! - Simple and economic fuel fabrication
- Inexpensive and compact fuel recycle I
- Improved Waste Management 1 - Smaller low-level waste volumes l - Inactivation of actinides J
e
- Deployment Flexibility and Low Cost Demonstration fI a
oenezuwces
- y su $=.
- q PRIS NUCLEAR STEAM SUPPLY SYST M Steam Reactor Generator Module Reactor Module Silo Silo CONTROL Sa m as
~
BOUNDARY e ROD DRIVES m aa , Commercial &s+ Nuclear Grade ,
.. n Grade I -
saw - -
e - r .-
- ,t 5 utsu -
9 ;
l Anta
( ' l ? ?,
w w.4 s
l' -c
. Hgja -
ld'lld
~
GUARD i VESSEL ~ "
- g. .
yM n a, o .. ui o.
+-- EM PUMP
- . ' . b
- 3 .
m b ..';Qa,n,t ,
- .a7,f _'
INTER-
. i ,
~#
- .P -
MEDIATE HEAT
- SA.
/ -
es arvon EXCHANGER i
- etmt -
.,g aracT e. ..
- y. k
.am ... Tte _
i
, , _ . REACTOR
/ ;, h ., g. , VESSEL
[,h((h !
CORE -
-j
,4%)
.t w
- 7,
.. i l Y\ f 873%os
"*N % AM b d 4** 'm" -
.w- - * -Ab- 4 e - m
% w ss W% M b db .e Mi &'h*#'BG=
i .
1 4
l O O O
! POTENTIAL ECONOMIC BENEFITS i
, FROM ADVANCED REACTORS i
q=_=----~=_= _2 - --- - _ . . =- -
=
1 i 1
100 -
1 N 90- Current LWRs l 80- ,
- ..., .,*, / - .
l 70- Current ::::: - ' ? '+ 4 + e m m ....f.' - /
j ,
md
- Coal !
- i . .M . d*.f.1-I:i
- 1 60- !
- i
~ 'J -- !': 4 .
i l
1987 Mills ;i:
. .. l l :1 Per kWh 50- I' NF - v c=m=_v=_
)
i l! }
" m. . -m:::'.v =_. wmm.=_m.
'l 4' 4 .. ?. ~ t
! jj 40-
~"'-" Gca.;.i'j:[m,,'.:~- . '],si3';{.,?ge.p;,_
l :: .p. , . : .,
}f . z, .s,.fg
'}
- 30- Advanced Reactor Systems '"-W d. ,L;.
- : x~. .
- ,i w/ Conventional BOP i
II 20-i ,1 10 -
I l l j O . . . . . . .
1 I O 200 400 600 800 1000 1200 1400 1' I l Plant Size MWe !
.1 8/29/88 ZM PEBTAR 1
1 l - _ _ _ _ _ _ _ _ _ _ - -__ ._ - . _
t
- O O O
-w, .
s __ -- -=_ - - - . - -
3 NUCLEAR PLANT COSTS CAN BE REDUCED g==-
1 LARGEST ,
1 UNCERTAINTIES !
y J YJ.t
'\ :
Equipment #
Field
& Q,\ S Labor
- Materials [Indirects !
Indirects Field Eppent ,
' Labor Mat rkis i
i i j CURRENT TECHNOLOGY ADVANCED REACTORS 1
tin> l arzsu m r
m__ _ _ _ _ _ . . - -__ - - . _ _ . _ -
O O O
- COMPARISON OF NUCLEAR PLANT L
COMMODITIES Q w . --.- - n- .L-._-------------a--
~ - lL'~" - ~.
% : 2".__
100 80 - - --------
if l; 60 -- - -
1 i Percent 40 -
Kfh'1_4M.I.E j EL-I-I4 X n- M l
20 0' O , , , .
1 Reinforcing Craft Structural Concrete Steel Steel (CuYd/ Labor (Tons / (Tons / MWe) (ManHr/
!( MWo) MWo) MWo)
.aw -
! O O O f l
._. y _ _ _ _ _
_ _ \
[ l i
2 2 +
L ADVANCED LMR PLANT COST TRENDS ,
i I t i :
\
) y=========~.--- -a.: -- .
= _
_w: -
-x j Simpler NSSS Requirements - - Lower Cost BOP l t 1.8 1 .! i
' MR A MWe)-
/ ~
I I
i 1.7 .
FOOL & LOOP 1978. l e, 1.6 -
+ @-
Q,l I*'
I 1.5 - - -
o -
1 .
3- l t l l #
'g LMR/ LWR 14 ~~ ~e.cnta -
o
- - ~ ~ ~~
h 0 Total Plant SPX-R epbca Cost Ratio- '
- 1.3 - - - 1N " "' - *.0$ g -
1 l2 S/KWe [- ;
{
, 1 <3 '
1.2
-+ - ;~
j os ts a-Loop s-J (Repbca-*3OO ; g+ j i ,j - MWe i 1.1 - - +
/ )1984 - 4 --
] j LSPB-POOL 1 , (9.phca-*3OO I } 1.0 ' - --
'#3*) '388 - --
+- -/ - - - - - -t lr CA u xuar Lua moo l l '
W'u<w.) 1986-87 i O.9 --
! 1 2 3 4 5 11 i i
8
j ?l i
- .w=- i I
t
O O O ADVANCED DESIGN INNOVATIVE FEATURE c- EXAMPLE- BOTTOM SUPPORT VERSUS - - - -
H I TOP SUPPORT REACTOR VESSEL ij Potential Bottom Support Cost Savings
( (% of Top Support Costs)
Operating Floor and Top Closure 49%
ii i i} Reactor Vessel 46% l i
i
- l[ Guard Vessel 73%
fj l!
ncept Sensitive Portions Only) 58% =
Savbgs b l; ti a !
ll
,. t
[
- Bechtel Corporation Analysis of Advanced LSPB MWe LMR Design tI l i
l i
~
o O O g== INTERNATIONAL COOPERATION _
ll 11
!!
- International Cooperation is Cost Effective Even Though j Advanced Reactor Concepts are Different . . .
- U.S. developing modular reactor, metal fuel cycle
[( - Japan and European Consortium developing monolithic 4 reactor, oxide fuel cycle ll li
- U.S. Completing Oxide Fuel R&D by Cooperating with
[ Overseas Partners E - Long-iife fuel demonstration tests with Japan L - Fuel transient testing with UK d - Improvements to aqueous reprocessing with Japan lt - Remote systems development with Japan, UK, France. FRG y - Clinch River /MONJU technology sharing with Japan il j
- Advanced Component Testing with Japan il y
- Advanced Component Design and Performance with a Japanese Private industries
i i
O O O l
_=__;, _ _ - .
ADVANCED LMR PROGRAM SCHEDULE l l l
- = .
-~ - =. .
x-- =
! l Oeri.II:n-Mel _ .
i :] IFR Fuel Cycle Oxido Fuel Cyclo l
- Fuel Development
- International Collaboration
! Ll i
- Core Design to Completo R&D _ i 0 . Safety Tests & Analysis
- Pyroprocess Development
! rj
- Rocyclo Demonstration !
!}
l l N ! E W W W !!! E R M 1 yl l {FY 1989l -
, __ - -.lFY1G91] . , - - - .
lFY 1993 l t! W
- e^- Wrvete.Sec kendlar_a dornational, interest I' I' ' FY 1987 FY 1988 FY 1990 FY 1992l ,
I onceptual Trao.-of f Advanced conceptual Design o.s ons . stud s D-4 +,t i t i
E:E IolalaIelteleymBDPI: Ire]elasI:lalE $ Select Reference Concept
. Advanced Steam Generator
{ elFR Technology g Decido Whether to Further Develop Reference Concept j i - Control and monitoring systems $ Defino Futuro Noods for Advanced Reactor Development and ,
, - Advanced materials Demonstration j ,i . - Other i
j a w eezu m l,
I
_m_ _ _ _ *A 4 a. d a.M._, a_ J.. _-A_m aL.__ _ . J. AAa _ , , _ . _ . ._ ._ m . __a_ - - -- - _ . - -_ z.a .*a ._A ._ m-__
l i
i
)
4 O
}
I i !
i J
l I !
t 1
- [
l 2 '
! 2 l 3 CD F- i l O W i i O l Q ,
! D g l n 2 !
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i b
'f i
1 I
I f
O :
f
, . o O O O IFR PROGRAM ESTIMATED FUNDING REQUIREMENTS g . _; _ - - .= ; . . .= --- _: = _ - _ __ . = _ - . . . ___=- - -
Year of Expendituro DO!!ars 125-100--
7 - Fuel Development 6 - Pyrop c Do srs 75 gy p--- ,.
- j Mdhons Q
- .7 O6 b 5 - Safety Tests
- I ~s5Q f _ , .- & Analyses
!! 50 - . '
4 - Core Design
, , _.;A- 3 - Fuel Cycle
"~-- Demonstration j
' ~
25 2 - Other f , , , -
f 1 - FaciEties
'{. .
(APE-W)
O Fiscal Year 1988 89 90 91 92 93 1994 Fud Devdopment 5.3 7.6 4.1 5.0 4.0 2.5 1.5 Pyroprocess Development 4.6 5.0 4.0 4.1 3.0 1.6 1.5 Safety Tests & Analyses 4.0 4.0 3.5 3.7 2.2 1.0 1.0 Core Design 1.8 4.3 1.4 1.7 1.2 0.6 0.5 Fuel Cycle Demonstration 3.8 5.3 2.0 2.0 1.0 0.4 0.5 Other 4.3 0.0 0.0 7.0 9.1 13.9 15.0 Facilities (AFL-W) 51.5 43.8 71.0 76.0 75.1 71.9 74.3 IFR + AFL-W Total 75.3 70.0 86.0 99.5 95.6 91.9 94.3 wwsezuwmo
O O O ANNUAL EXPENDITURES FOR CIVILIAbi i NUCLEAR R&D p-w . -- -- .
] 1.400 - !
1
) '
Gas Reactors
! 1.200-- ^s-m#
N
. / ,. / N .
s '
LWR i
1
\
1.000-.
/ l.. /' 1 ' .. . ..,., \ x l ~'
Millions Water Cooled Breeder
)
'N N j
l I Ii 4
1987 of 800 - /[/ ,
\ CMR s
,,,~ ~. \
I
'l Dollars f 's -h, i j
g .,- ,
~~-
j ,j 600 - g ,\
LMR w .-
i i
l j; 400 -
N '(.\ N l
I N ., \
l r 200-
- % .% ' %:= -
i
-1 0- . i , i , i i i , i i , , , i 1973 74 75 76* 77 78 79 80 81 82 83 84 85 86 87 88 1989
- 15 Months Fiscal Year !
.m-m um
i O O O DRAFT SER ON PRISM 1
\
Presented to the ACRS Advanced Reactor Subcommittee 1
October 5,1988 .
I Thomas L. King, Chief '
ARGIB RES
O O O l
PURPOSE OF PRESENTATION
.i i
- To summarize the staff review and conclusions resulting from our review of the PRISM conceptual design i
j
- Information briefing only i
- ACRS Letter requested on PRISM conceptual design and staff's Draf t SER af ter November Full Committee meeting l
- l l
l l
l l
l f
i
O O O SER STATUS l
- ARGIB REVIEW OF FRISM COMPLETE
- SOME SER SECTIONS NEED UPDATING:
I - Chapter 3 - Criteria
- Chapter 6 - Containment
- - Chapter 15 - Safety Analysis
- COMPLETION OF INTERNAL REVIEWS:
! - RES Management 10/88 j - ACRS Full Committee 11/ 8 8
- NRR 11/ 8 8
- CRGR 11/ 8 8
- To Commission 11/ 8 8 1
i i
O O O BACKGROUND '
i 1
- Formal staff review of PRISM started in December 1986
!
- Main documents submitted by DOE for review:
l - Preliminary Safety Information Document (PSID) l - Probabilistic Risk Assessment (PRA)
- Formal NRC/ DOE meetings - April 1987 - November 1987
- Followup meetings in January 1988 and July 1988
- Eleven Amendments to the PSID issued by DOE
by DOE /NRC on the PRISM design and staff review:
-Subcommittee 2/25/86 l 6/17/87 i 8/25-26/87 at EBR-il i -Full Committee 4/10/8 6 l 2 /11/ 8 8 ;
- Draf t SER provided to the Subcommittee on 9/10/88 i
t
O O O PURPOSE OF STAFF REVIEW OF PRISM
\
i
- Provide guidance early in the design process on the acceptability /licensability of the design, including:
I
- Licensing criteria
- Potential of the design to meet those criteria I - Acceptability of proposed supporting R & D programs i
. Basic guidance to the staff on conducting the review was provided by the Commission's Advanced Reactor Policy Statement (July 1986):
- Perform
- Earliest possible interaction" with
- applicant,
! vendors, and government agencies"
- Recognize reactors that "Utilize simplified, inherent or other innovative means to accomplish their functions" (i.e., review designs on their own merits)
- As a minimum, advanced reactors should provide at least the same degree of protection of the public and the environment that is required for current generation LWRs.
Expect enhanced safety.
1 I l
O O O STAFF REVIEW PROCESS
- Described in SECY 86-368, dated 12/10/86, for the I
three DOE concepts
- NUREG-1226 documents general review process:
- Definition of Advanced Reactors
- Information which should be submitted for staff review l - Review approach:
- Build upon applicable LWR regulations and criteria
!
- Utilize Commission's Policy Statements (Safety Goal.
Severe Accident, Standardization) as guidance in
, deve!nping criteria and assessing unique attributes I
of advanced designs
- Assess supporting R & D needs, including need for prototype a
l j
4
O O O
SUMMARY
OF SER I
- Pre-Application SER - should be recognized as preliminary guidance, not a statement of design approval. It is a limited assessment of design features and research and j testing needs
- Overall Conclusion - provided certain fundamental design l
- issues can be resolved, the staff can conclude l that the PRISM design has the potential to achieve a level q
of safety equivalent to current generation LWRs :
4 - response to certain bounding events leads to fuel melt
! and/or Na boiling (i.e., potential for positive reactivity I feedback accident / core melt accident) i -adequacy of containment i
)
- POTENTIAL FOR DESIGN SOLUTIONS l
4 i
i
O O
SUMMARY
OF SER
- continued I
- Final Determination of PRISM Acceptability - contingent on the following :
1 1). Satistactory resolution of the issues identified in the staf f's SER, including completion of the analysis, j
research and testing, and Commission action on the Four Key Licensing issues 2). Completion of final design und licensing review by NRC 3). Successful design, construction, testing and operation i
of a prototype reactor prior to Design Certification i
l I
i l
I i
e a q
O O O FAVORABLE CHARACTERISTICS
- The design has many favorable characteristics 1). The potential for only minor core damage and fission product release over a wide range of severe challenges to the plant 2). Ihe reduced dependence on human actions and the reduced vulnerability to human error 3). The long response time of the reactor under many accident conditions, providing time for evaluation and corrective action 4). The capability to demonstrate by test the significant safety features and performance of the plant over a wide range of events
i O O O Power Reactor Inherently Safe Module
]
Safety Evaluation Report i
j Presented to the ACRS
! Advanced Reactor Subcommittee
! October 5,1988 i
Ralph R. Landry ARGIB j RES l
1 i
l
O O O WORLD LMRs REACT-OR THERMAL POWER POOL / LOOP CRITICAL COUNTRY EBR-Il 62.5 MW Pool 1963 U.S.
Phenix 563 MW Pool 1973 France PFR 559 MW Pool 1974 U.K.
FFTF 400 MW Loop 1980 U.S.
SNR 300 736 MW Loop ----
FRG Monju 714 MW Loop ----
Japan ~
Superphenix 3000 MW Pool 1985 France
O O O PRISM PLANT PARAMETERS Overall Plant Number of Reactor Modules 9 Plant Thermal Power, MWt 3825 Net Electrical Power, MWe 1245 Number of Control Rooms 1 Reactor Module Core Power, MWt 425 Primary Sodium inlet, F 610 Primary Sodium Flowrate, GPM 40 800 Int. Sodium Inlet, F 540 Int. Sodium Flowrate, GPM 41 000 Steam Cycle Saturated
I O O O CHAPTER 1 DESIGN REVIEW
- APPROACH BASED ON Advanced Reactor Policy Statement & NUREG 1226 Severe Accident Policy Statement (NUREG 1070)
Safety Goal Policy Statement (NUREG 0880, Rev 1)
Standardization Policy Statement (SECY 87-193)
- CONSIDERED Conservative engineering design practices?
Redundancy and diversity?
Deterministic and probabilistic analyses?
Defense in depth?
O O O CHAPTER 2 SITE
- GESSAR 11 SITE Exception: Larger snow load
- ENVELOPE 90% OF CURRENT SITES
- SAFE SHUTDOWN EARTHOUAKE Zero period acceleration of 0.3 g Reg Guide 1.60 response spectra scaled to ZPA Motion applied at ground surface 4
l l
i
)
i i
l
O O O l CHAPTER 3 DEFENSE IN DEPTH I
- PREVENTION l Conservative design assumptions j
Appropriate codes and standards l
Quality in derign, construction and maintenance
- PROTECTION Reliability Redundancy / diversity
- MITIGATION Delay / limit release of fission products I
- EMERGENCY PLANNING Protection of the public I
l l
l l
. - . -. ._ . - .- _ .,=
~
1
- O O O l CHAPTER 3 GENERAL DESIGN CRITEP.lA l
1 l
- AGREE WITH
! 35 GDCs i
i
. TAKE EXCEPTION TO 6 GDCs j
- NOT APPLICABLE 9 GDCs
! = STAFF CONSIDERING MODIFICATION 5GDCs I
1 4
i l
1
)
1
O O O CHAPTER 3 BOUNDING EVENTS
- THESE EVENTS ARE INTENDED TO BOUND THE LMR DBA AND BDBA SPECTRUM TO ACCOUNT FOR PRA UNCERTAINTIES. THEY ARE EXPECTED TO PROVIDE CONSERVATISM IN SELECTING A SSST.
- INADVERTENT WITHDRAWAL OF ALL CONTROL RODS Without scram for 36 hrs With forced cooling With only RVACS
- STATION BLACKOUT - 36 hrs
- LOSS OF FORCED COOLING PLUS RACS/RVACS With 25% RVACS unblockage after 36 hrs
- INSTANTANEOUS LOSS OF FLOW FROM ONE PUMP Coastdown of other pumps
- STEAM GENERATOR TUBE RUPTURE Justifiable number of tubes, sequence to be defined Failure to isolate or dump water from SG Without forced cooling
O O O CHAPTER 3 BEs CONT'D
- LARGE Na LEAK MEFS break of IHTS pipe Reactor vessel leak (critical size)
- EXTERNAL EVENTS Events to be defined consistent with those imposed on LWRs
= FLOV/ BLOCKAGE OF ONE FUEL ASSEMBLY I
j i
l I
O O O CHAPTER 4 i
REACTOR l
1
)
i
- CORE MATERIAL
) Ternary alloy: U-26Pu-10Zr
! Blanket alloy: U-10Zr
! Clad: HT-9 i
- CORE CONFIGURATION Fuel assemblies 42 Fuel rods per assembly 331 l
- Internal blankets 25 Radial blankets 36 Reflector assemblies 42
! Radial shields 48
! Control assemblies 6 1
l
- FUEL LIFE 1 Refueling intervals, mon 20
) Fuel life, mon 60 Peak burnup, mwd /kg 147
)
l
O O O CHAPTER 4 CRITERIA
- FUEL IS NOT DAMAGED BY Normal operation Anticipated operational occurrences
- NO FUEL MELTING
- NO SODIUM VOIDING
- FUEL DAMAGE WILL NOT PREVENT CONTROL ROD INSERTION
- NUMBER OF FAILURES IS NOT UNDERESTIMATED
- COOLABILITY IS ALWAYS MAINTAINED 1
l l
i e
4 l
,,n . - _ , -. ,7-. - . .
O O O CHAPTER 4 FUEL CYCLE
- PRISM PLANT: INTEGRAL FUEL REPROCESSING Fabrication Reprocessing Waste management
- REPROCESS BY PYROMETALLURGICAL TECHNIQUE Laboratory scale demonstration
- COMPOSITION OF REPROCESSED FUEL YET TO BE DETERMINED l
O O O CHAPTER 4 CLAD MATERIAL
- HT-9 STEEL Low swelling with irradiation Relatively new - little data
-Thermomechanical properties
-Batch to batch consistency
-Welding-fabrication techniques a CONCERN: NEED FOR DATA BASE
- EXPECT: R & D EFFORT TO SUPPLY DATA BASE
O O O CHAPTER 4 FUEL PERFORMANCE
- FUEL LIFE LIMIT Creep rupture of HT-9 cladding from plenum pressure due to fission gas
- FAVORA3LE Highly porous fuel High cladding strength Fuel-coolant compat'.bility
- UNDER INVESTIGATION Local ef fects along pin axis, e.g., fission gas retention and swelling Species migration with burnup
- CONCERNS Low fuel / clad eutectic temperature Fuel restructuring Sodium bond / porosity behavior l
O O O
- CHAPTER 4 REACTIVITY CONTROL i
- SIX ACTIVE CONTROL RODS Safety grade trip Latch system similar to CRBR design Drive-in follows trip - 2 000 lbs force
- l
.i i
- CONCERN l Single active system i
e I
i
O O O CHAPTER 4
- INHERENCY-REACTIVITY COEFFICIENTS
- DOPPLER Smaller negative factor than for oxide fuel
- SODIUM DENSITY Positive coefficient with voiding l
- AXIAL EXPANSION Negative coefficient - function of burnup
- RADIAL EXPANSION l Negative coefficient - slow acting
. BOWING i Limited free bow to keep coefficient negative
. CR DRIVELINE EXPANSION j Expansion inserts control rods l . VESSEL EXPANSION Withdraws control rods i
- l
O O O CHAPTER 4 REACTIVITY COEFFICIENTS
- ESTIMATES Consistent with designer's
- CONCERN Relatively high uncertainties (10-25%)
Fast flow reduction leads to sodium voiding
- FAVORABLE Prototype test program
- ISSUES is inherency a shutdown "system"?
Can we accept a positive void coefficient?
_ ._ _ -= _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _
O O o R&DISSUES CHAPTER 4
- FUEL Low eutectic temperature Sodium logging Thermal properties with burnup Cladding properties Slow overpower behavior e FUEL FAILURE PROPAGATION /RBCB
- REACTOR PHYSICS Reactivity feedback confirmation
4 O O O CHAPTER 5 HEAT TRANSPORT PRIMARY INTERMEDIATE WATER 1
l i
IHX SG i
l I
l J
i l o o o I CHAPTER 5 MATERIALS
)
- LARGE BELLOWS With 30 in pipe l
- GIMBALLED JOINTS
- RUPTURE DISKS I 30 in diameter l
- HT 9 STEEL l Lack of data on properties
! Wide thickness range - 22 mils to 4 inches a
5 I
p l
I 1
o O O DESIGN ISSUES i
CHAPTER 5 t
{
- EM PUMPS l
Seismic isolation and separation of synchronous machines 11
'l j
i 4
i l
l l
l l
l 0 o o
- R&D ISSUES 4
CHAPTER 5 i
- RVACS
) Assess improvements due to RV/CV gap filled with sodium
. STEAM GENERATORS j Double wall tube thermal performance l
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i O O O
- DESIGN ISSUES CHAPTER 6 1
- CONTAINMENT Acceptability of design without conventional containment
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e e e DESIGN ISSUFS 1 CHAPTER 7 l
- RPS ISOLATION Requirements needed for numerous 9PS/PCS interfaces
- EM PUMPS Temperature beyond current experience base l
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O O O DESIGN ISSUES CHAPTER 13
. EMERGENCY PREPAREDNESS Acceptability of ad hoc planning - Commission decision
. SECURITY No communications described Adequate sabotage protection
. ROLE OF OPERATOR One operator for multiple units Non-safety function of operator
. CONTROL ROOM Safety classification Remote shutdown area
! O O O PRISM SER l CONCLUSIONS Provided certain fundamental design issues can be resolved, i
the staff can conclude that the PRISM design has the l potential to achieve a level of safety equivalent to current l generation LWRs. These fundamental issues are:
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- 2. Adequacy of containment i
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O O O PRISM SER
- FAVORABLE FACTORS l
- Passive means of reactivity shutdown and decay heat removal are avaliable l
- Potential for only minor core damage and fission product release for many events ,
- Reduced dependence on human actions and reduced vulnerability to human error
- Long response time under many accident conditions
- Capability to demonstrate by test significant safety i features and plant performance i
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O O o PRISM SER
! CONCLUSIONS CONTINGENCIES -
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- BEYOND CONCEPTUAL DESIGN, ACCEPTABILITY DEPENDS ON:
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- Satisfactory reeclution of all issues identified i
i a Completion of R & D
= Completion of final design and licensing review
- Successful design, construction, testing and operation of a prototype reactor i
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