ML20211B730

Transcript of Commission 861009 Briefing in Washington,Dc Re Advanced Reactor Designs.Pp 1-73.Related Documentation Encl

ML20211B730
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Issue date:	10/09/1986
From:	NRC COMMISSION (OCM)
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REF-10CFR9.7 NUDOCS 8610210075
Download: ML20211B730 (267)

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0RIGINAL UNITED STATES OF AMENICA NUCLEAR REGULATORY COMMISSION l

In the matter of:

COMMISSION MEETING Briefing on Advanced Reactor Designs (Public Meeting)

Docket No.

Location: Washington, D. C.

Date: Thursday, October 9, 1986 Pages: 1 - 73 ANN RILEY & ASSOCIATES Court Reporters 1625 I St., N.W.

,. Suite 921 Washington, D.C. 20006 j (202) 293-3950

, 8610210075 861009 PDR 10CFR

( PT9.7 PDR L

(O) v 1 O I SC LA I MER 2

3 4

5 6 This is an unofficial transcript of a meeting of the 7 United States Nuclear Regulatory Commission held on S 10/09/86 . In the Commission's office at 1717 H Street, 9- ' N . tJ . , (Jash i ng t on , D.C. The meeting was open to public 10 attendance and observation. This transcript has not been 11 reviewed, corrected, or edited, and it may.contain 12 inaccuracies.

! D l

13 The transcript is intended solely for general l

14 informational purposes. As provided by 10 CFR 9.103, it is 15 not part of the formal or informal record of decision of the 16 matters discussed. Expressions of opinion in this transcript 17 do not necessarily reflect final determination or beliefs. No 18 pleading or other paper may be fi1ed with the Commission in 19 any proceeding as the result of or addressed to any statement 20 or argument contained herein, except as the Commission may i

, '21 authorize.

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A 1

1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 ***

4 BRIEFING ON ADVANCED REACTOR DESIGNS 5 ***

6 PUBLIC MEETING 7 ***

8 Nuclear Regulatory Commission 9 Room 1130 10 1717 H Street, Northwest 11 Washington, D.C.

12 13 Thursday, October 9, 1986

/ 14 15 The Commission met in open session, pursuant to 16 . notice, at 10:04 a.m., the Honorable LANDO W. ZECH, JR.,

17 Chairman of the Commission, presiding.

18 COMMISSIONERS PRESENT:

3 19 LANDO W. ZECH, JR., Chairman of the Commission i

20 THOMAS M. ROBERTS, Member of the Commission 21 JAMES K. ASSELSTINE, Member of the Commission j 22 FREDERICK M. BERNTHAL, Member of the Commission l

23 STAFF AND PRESENTERS SEATED AT THE COMMISSION TABLE:

24 PAUL BOLLWERK 25 SAMUEL J. CHILK

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i 2

i 1 A. DAVID ROSSIN

{

JOHN S. MCDONALD 2

3 ROBERT T. LANCET

. 1 l 4 CHARLES L. STORRS l

! 5 J.S. ARMIJO 6 GLENN SHERWOOD

) 7 NEIL BROWN 8 T.E. NORTHUP i 9 DANIEL MEARS b

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1 _P R O C E E D'I N'G S 2 CHAIRMAN ZECH: Good morning.: Commissioner Carr 3 will not.be with us this morning. The Commission is meeting 4 today to hear from the Department of Energy and others on the 5 status of the advanced reactors concept being supported by the i 6 Department of Energy.

7 This is an information briefing-and the Commission 8 does not anticipate any formal decision or action as a result 9 of today's meeting, other than to consider the material 10 presented. The conceptual designs that are to be discussed i 11 today are reflective of the concepts that the Commission was 12 considering when the advanced reactor policy statement was I

13 developed.

! N 14 We have a rather ambitious task this morning. I'd

15 appreciate our presenters moving along into the briefing, and 16 I'd ask my fellow commissioners to limit their questions to 17 the degree that they feel they can for clarification. I know l 18 we're going to be afforded ample opportunity to address in 19 much greater detail many of the design considerations as we

( 20 proceed. I have an appointment that I must keep later this 21 morning and I trust we'll complete our meeting by 11:45. If 22 not, I must leave at that time.

23 Do any of my fellow commissioners have any opening 24 remarks?

25 COMMISSIONER ASSELSTINE: No.

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4

'1 CHAIRMAN ZECH: Dr. Rossin,' .welcome and 2 congratulations on your new assignment in the Department of 3 Energy. I believe this is the first time you've been before 4 us in your new role. We're pleased to have you here this 5 morning and please proceed.

6 MR. ROSSIN: Thank you, Mr. Chairman, you just took 7 the first line of my speech.

8 As Assistant Secretary for Nuclear Energy I'm I

9 responsible for DOE's advanced reactor development program.

10 You will hear presentations ~by three leading organizations 11 which are working on new innovative designs for nuclear 12 reactors. I want to thank the Commission for its support in 13 its licensing review of these designs. As I will say and 14 others will say, this is absolutely crucial if we're going to l 15 make any progress in this area.

16 These reactor power plants are smaller than the

( 17 sizes found in current generating stations. DOE is supporting l 18 these efforts in the hope that a decade from now 19 U.S. utilities will have th'a option of ordering an alternative 20 reactor generating unit based on one or more of these 21 concepts. To realize this goal several positive developments 22 are needed.

l l

23 First, joint industry-DOE support of design and 24 development work. Finding a way to finance the first of a 25 kind unit. Obtaining a license from NRC for construction and 1

5 1 operation of the first unit. The co'nfidence that the 2 political conditions will permit such a project to proceed.

1- 3 Why should the government permit an alternative d

4 design at this time? Well, let me throw out a few ideas not 5 in any definitive order. Utilities want a competitive a

6 marketplace. There are national security advantages to 7 diversity among energy supply sources. There is little 8 likelihood that-the private sector alone could justify the 9 investment required considering the many years before return 10 on investment could be realized. The people of the U.S. have 11 the right to expect their government to take the lead in long j 12 term programs aimed at assuring a better life for their 13 children and their grandchildren.

1 14 One more point. The lessons of what I call the

, 15 environmental decade have been learned by responsible 16 citizens. We cannot ignore the future and live for the 17 present only. We must conserve our resources and protect the 18 environment for the future.

19 Will the market buy smaller modular reactors? I 20 don't know the answer yet. The critical factors are known 21 however. Today's plants require huge capital investment and 22 long construction times. So while economies of scale are 23 still real for power plants, and especially for nuclear power 24 plants, these economies can be lost due to delays and 25 uncertainties, i

6 1 The economies of smaller pl' ants can be realized with 2 shp fabrication of as much as practical, better and more 3 efficient production and quality _ control, and economies from 4 licensing and production of repeat units. To achieve these 5 attractive marketing objectives, the design must satisfy b

6 regulatory requirements and be certified to be safe enough to 7 build and operate. This is a prerequisite to private l

8 investment in a factory-type production facility.

t 9 -I hope to see NRC develop a licensing mechanism that 10 offers safety certification of a design. The uncertainties 11 that have characterized the past decada must be removed in 12 order for these new plants to enter the marketplace. In my 13 personal view, I believe that before a modular unit will be 14 ordered there must be confirmed evidence that new nuclear i 15 power plants can be ordered, built and operated. It is my 16 hope that a simpler, advanced light water reactor can be

, 17 ordered, licensed and built without a prototype or a demo 18 plant and the design can be certified so that others can be 19 ordered.

20 It's really not a question of advanced light water 21 reactor or advanced small or modular reactors. The first is 22 likely to be a prerequisite to the second. I would say more 23 than likely. The first, in my view, is essential to order 24 reestablish nuclear power in the marketplace. And the key is 25 the establishment of licensability or certification, something l

7 1 that assures regulatory acceptance. '

b

\ 2 I'm pleased this morning to introduce three 3 presentations by three different reactor designers: Rockwell 4 International, GE, and GA Technologies. I'm going to ask them 5 to come to the table individually, make their presentations i

6 and respond to your questions. I'm willing to respond to any 7 preliminary questions you wish and I'll be available for your 8 questions throughout the morning. Thank you.

4 9 CHAIRMAN ZECH: Thank you very much. Proceed.

10 COMMISSIONER BERNTHAL

I guess you've just answered i

11 my question, you will be available, because I had some 12 questions I wanted to address to DOE. But it would be better, 13 I think to hear the presentations first.

14 MR. ROSSIN: .That's fine with me, Commissioner, I'll 15 be here.

16 CHAIRMAN ZECH: Fine, let's do that and then we'll 17 have questions later.

18 MR. MCDONALD: Mr. Chairman, I'm accompanied this

19 morning by Bob Lancet, who's the manager for Rockwell safety 20 and licensing, and Dr. Chuck Storrs of Combustion Engineering.

21 CHAIRMAN ZECH: Welcome and proceed.

I 22 MR. MCDONALD: Thank you. Could we have the first

23 slide, please?

24 (Slide.]

25 MR. MCDONALD: We really appreciate the opportunity l

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, 1 to address you this morning on what:w's refer to as the Sodium 2 Advanced Fast Reactor program, and we use the acronym SAFR.

t 3 In the time that's available, I'd like to introduce you to our I

4 4 program approach, present a brief overview of the design. I l 5 think from the design it'll be apparent from the simple and l

j 6 self-protective features that our concept can provide a high 1

7 level of safety. We'll touch on our economic projections and j 8 finally outline the approach we recommend for the first plant.

! 9 Can we have the next slide, please?

10 (Slide.) 1 Y

11 MR. MCDONALD: Several years ago prior to the i

I 12 initiation of the DOE innovative design program our centractor i

13 team, Combustion, Bechtel and ourselves with significant 14 assistance from Argonne National Laboratory took a hard look i

l 15 at'the complex of concerns that were confronting nuclear

! 16 power.

i 17 And we arrived at the general SAFR concept is one i I

18 having a really significant chance of being able to alleviate 19 many of the nuclear plant concerns. And the design I'll be 20 describing for you, we believe, can provide natural safety --

t 21 I'll elaborate on what I mean by that as we go along --

22 competitive economics, much shorter construction schedules 23 than have been recently experienced. That will pay off in 24 terms of reduced interest during construction, reduced 25 risks in utility planning. And finally, a user friendly plant

9 1 will improve availability.

2 Can we he the next one, please?

3 [ Slide.]

4 MR. MCDONALD: I expected that some of you might not 5 be familiar with the properties of liquid metal reactors, j 6 liquid metal reactor technology, so I took the liberty of 1 7 offering'for just a moment a few fundamentals about sodium 8 physical properties and how they can translate into very 9 advantageous plant characteristics.

t 10 The thermal conductivity of sodium, for example, is 1

l 11 higher than that of water by a factor of about 40. Sodium has 12 a very low vapor pressure. At all the temperatures that the i

13 sodium will experience in a liquid metal reactor, the vapor 14 pressure will be well below atmospheric. And what this allows 4

15 you to do is operate with a reactor vessel at near j 16 atmospheric, just has to carry the hydrostatic head of the 17 sodium in the reactor assembly. The secondary loop is at l

18 about 100 psi, pounds per square inch. And this compares with i 19 pressures in a light water reactor of 1500 to 2500 psi, that 20 general range.

i

! 21 Sodium is non-corrosive of the stainless steel 22 structural materials that are used in these plants. It has a I

23 very hard boiling point. At atmospheric pressure sodium boils

! 24 above 1600 degrees Fahrenheit, so this is several hundred 25 degrees higher than any temperature that you'll experience 1

10 1 during operations. Has very strong natural convective

)

2 tendencies.

i 3 So collectively, what these wind up doing is 4 allowing you to design a fairly efficient system, low pressure 5 coolant system. You can achieve high plant efficiencies 6 approaching fossil plant efficiencies. Use conventional steam 7 equipment. And there are an abundance of safety J 8 characteristics that I'll enlarge on later on.

9 COMMISSIONER BERNTHAL

One quick question.

4 10 MR. MCDONALD: Yes, sir.

1

!' 11 COMMISSIONER BERNTHAL: What makes something have 1

12 strong natural convection? Is that the coefficient of l 13 expansion or something?

14 MR. MCDONALD: Yes, the way in which the sodium 15 density depends on temperature differences.

, 16 COMMISSIONER BERNTHAL: Fine, okay.

{ 17 MR. MCDONALD: I did want to touch on this point 18 which is very important to people who are managing plant 19 operations and maintenance, the worldwide experience with 20 liquid metal reactors is that the occupational dose rates 1

21 experienced by personnel, operating personnel, are 22 substantially lower with liquid metal reactors than with other j 23 types of. reactors. Lower by a factor from 10 to 50 depending

! 24 upon what particular conditions you're comparing.

25 Can we have the next slide, please?

)

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11 1 (Slide.] '

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( 2 COMMISSIONER ASSELSTINE: John, in terms of design 3 characteristics, would this design have a positive void 4 coefficient?

5 MR. MCDONALD: Yes, it does. Something of less than 6 two dollars and something. Very small, and we can elaborate 7 on a little bit later on.

8 COMMISSIONER ASSELSTINE: Okay.

9 MR. MCDONALD: 'This chart shows our contractor team

'l 10 that's carrying out the SAFR program for the Department of 11 Energy. Rockwell, Combustion and Bechtel have been i 12 collaborating for about 10 years now on trying to develop an i 13 improved liquid metal reactor designs. And we're really

\ 14 pretty pleased about the way the SAFR design is taking shape.

15 Below the line down here are a number of I

l 16 organizations that are active on the DOE base program. And l

17 listed are a number of different tasks which are specifically 18 designed to support our SAFR effort. These are all being 19 carried out under an integrated work plan, and Argonne in 20 particular is doing some very helpful work for us on metal 21 fuel.

22 The point I wanted to make was that they're really 23 substantial resources being focused into aid and developing 24 and to develop and substantiate the SAFR design.

25 Next chart, please.

l

12 (Slide.]

1 O)

\, 2 COMMISSIONER ROBERTS: Pardon me, would you give me 3 those acronyms below that.

4 MR. MCDONALD: Can we go back to the previous chart?

5 (Slide.]

6 This is Argonne National Laboratory. This is the 7 Energy Technology Engineering Center, which is operated by 8 Rockwell for DOE. This is the Hanford Engineering Development 9 Laboratory, which is located in the Hanford Reservation in 10 Washington.

11 Los Alamos National Laboratory, Oak Ridge National 12 Laboratory. This is Atomics International, a segment of 13 Rockwell. And this is the Westinghouse Division in Walks 14 Mill.

15 The next one, please?

16 (Slide.]

17 MR. McD0NALD: Dr. Rossin outlined the objectives of 18 the Innovative Design Program. This chart just shows the 19 logic that we're following in achieving those objectives, 20 attempting to achieve those objectives.

i 21 During this 37 month design period, and we're at 22 about the halfway point right now, we're in the procesc of 23 trying to tune and tailor and prove our design in such a way 24 that we can, at the end, have a design that best meets user 25 needs. And as we move along, we're factoring in the data that

13 1 flows in from the base program results, factoring in feedback 2 from our utility interactions.

3 And let me digress just long enough to say that we 4 have met with a large number of utilities and are continuing 5 to, on a regular basis. Individual meetings in their offices, 6 seminars periodically. We've done some joint studies with 7 them from time to time. They participate in our design 8 reviews and we're getting some very beneficial help in 9 identifying concerns and changes we should make to accommodate 10 owner and operator interests and concerns.

11 I've got to confess that while they're all very 12 interested and supportive of the concept, none has been ready 13 yet to step up and take a major financial role. We hope and

\x 14 trust that that will change as the program goes along and as

15. we get better evidence on the licensability of the safest.

16 We had a very beneficial dialogue with your staff so 17 far and that has certainly been influential in helping us to l

18 improve the design.

19 And then finally, to aid us in addressing the 20 institutional concerns and to formulate a strategy for 21 commercial introduction of this concept, the company expects i

22 to retain what we refer to as the institutional advisory 23 committee. And here we have a dozen men who are the leaders 24 in the various areas, such as finance, public acceptance, and 25 regulation and user utility needs, management.

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14 1 These gentleman have been m'onitoring the program 2 carefully. They're very enthusiastic. They're helping us 3 work out some of the financing arrangements that I'll outline 4 later on. As an outcome of this overall effort, we expect 5 that we will be able to produce something that we believe will 6 be an attractive option for future use by utilities and it 7 will be accompanied by a number of plans which will provide a 8 solid basis for moving ahead on a prototype program.

9 The next one, please?

10 (Slide.]

11 COMMISSIONER ASSELSTINE: John, at some point do you 12 plan to address the licensing approach that you had in mind, 13 the extent to which you think your design is licensable under k/

m 14 existing regulations and what kinds of changes you have in 15 mind?

16 MR. MCDONALD: We'll touch on that.

17 COMMISSIONER ASSELSTINE: Okay.

18 MR. MCDONALD: If we can go to the next one, I'd 19 like to just highlight the design for you for a moment. I'd i

20 like to lead into that, though, by just indicating that we did 21 establish these goals early, based on that assessment of l 22 market needs that I referred to earlier.

i 23 And a number one priority was safer use to assure 24 public protection and to provide a safe plant without reliance 25 on the need for fast operator action. And we believe we can

J 15 1 show that the risk will be sufficiently low that we can N

2 consider eliminating the need for evacuation and we would 3 accomplish that.

4 We are accomplishing it in the design by 5 incorporating features such that by the actions of natural 6 forces, such as gravity, the plant is able to on its own 7 establish and maintain safe conditions following any event.

8 We'll touch further on some of those potential initiating 9 events later on.

10 Under this umbrella, the safety blanket, then we are 11 able, we think, to provide enhanced investment for the owner's 12 investment. We certainly are going to have to show g- 13 competitive credible energy costs and this has been our 14 target, and we think we do well compared to that.

15 A very sharp construction schedule, which of course 16 will reduce costs and improve predictability of construction 17 schedule, reduce the risk.

18 other factors, I won't dwell on.

19 Can we go to the next one, please?

20 (Slide.]

21 COMMISSIONER ASSELSTINE: John, in terms of your 22 cost estimate, is that target what you expect to cultivate say 23 in the 1990s?

24 MR. MCDONALD: I'll have a chart a little later on 25 that will show what our projections of coal and others.

\

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16 1 COMMISSIONER ASSELSTINE: And could you also address

( j 2 the extent to which that cost target is influencing the safety 3 of the design?

4 MR. MCDONALD: Okay. Why don't you make a note of 5 that, Vince.

6 COMMISSIONER ASSELSTINE: In terms of things like 7 containment, power density, those kinds of things.

l 8 MR. MCDONALD: Okay. I won't belabor these goals.

9 I think they're all important. I guess a keystone is that --

10 our aim is to obtain certification for a standard plant 11 design. And that's really critical to our achieving many of 12 the other goals that we've laid out here, the ability to

, 13 replicate a standard design and get the cost and schedule 14 benefits to go with that, is very important with the effort.

15 Our design, we feel at this stage can meet all of i

16 these goals. In fact, our evaluation indicates that we can 17 exceed them in several areas.

18 Go on to the next one, please?

19 (Slide.)

20 MR. MCDONALD: One last chart before describing the 21 design. Our -- again, very early before the Innovative Design 22 Program began, our contact became assisted by Argonne National 23 Laboratory and Chicago Bridge and Iron. We set out very early 24 to try to find an answer to the question -- an honest, 25 objective answer to the question what approach really has the l

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17 1 best chance of achieving the goals that we set for ourselves.

2 And we made a-very comprehensive evaluation over 3 about a nine to ten month period. There were numerous 4 cross-checks. We recognize the vulnerability of any kind of 5 assessment like this. And we all pretty well satisfied 6 ourselves that for the environment we have and the goals we 7 set, the characteristics shown in this box probably are the 8 best choice. And as we've gone along we've -- I think --

9 convinced ourselves more and more that that's the case.

10 Let me focus just a minute on a size question. In 11 our evaluations, we looked at plants that -- monolithic plants 12 that can be steel fabricated in the size range of 1,000 to 13 1,350 megawatts. And then looking at the family that can be O 14 fabricated in the factory and then shipped to the site either

15 by rail or barge, we looked at module sizes ranging from 110 16 up to 350.

17 These can be arrayed together as necessary, to 18 provide the generating capacity that's needed at a particular 19 station.

20 We settled in on this size because it yields the 21 benefits, the cost, the schedule, the quality benefits that go 22 with a factory fabrication environment. At the same time, it I

23 yields a reduction in fuel construction labor and quality 24 assurance activities, a reduction in field' construction 25 schedules, all of which reduce costs and risks.

O.

I

18 1 We also -- another reason for that particular size 2 .is this is still small enough to enable you to.get the 3 benefits of redundant diverse passive decay heat removal 4 systems, drawing heat directly from the reactor assembly, as 5 well as other benefits to go with^a smaller reactor size in 6 the safety area.

7 Additionally, as far as the factory envelope is i 8 concerned, you get econcmies of scale with a 350 megawatt 9 size, compared to smaller units. I think on the surface of 10 it, one can anticipate some economies of a 350 megawatt size 11 compared with the others, to supply a given level of power.

i 12 For example, it would take about three times as many of these 13 smaller units as it would the 350 megawatt size.

14 And that means you have to identify and machine and i 15 fit up and weld and clean and inspect three times as many 16 parts, three times as many instruments, three times as much 17 wiring,-three times as many reactors to operate. So just

'18 on the surface, I think one would anticipate the result that 19 we came up with, which is 350 megawatt is the winner.

20 COMMISSIONER BERNTHAL: So that's basically the 21 biggest factory fabricatable plant that you could come up 22 with? Isn't that the criteria?

23 MR. MCDONALD: We might be able to squeeze out more, 24 but this is about the biggest, consistent with being able to 25 provide these redundant diverse passive decay heat removal 1

19-

) 1 systems I'll describe later on, which draw energy directly 2 from the reactor assembly witho'It having to go through much 4

3 intermediate piping system.

4 We've gone with the pool configuration, which seems 5 to be the emerging choice around the world, and we're using as 6 a reference metal fuel because Argonne is doing a super job 7 in developing that fuel. It does offer fuel practical cost 8 advantages over the oxide. It has some safety advantages over 9 the oxide.

, 10 However, they're retaining the ability to use oxide 1

11 as a fall back, if that becomes necessary for a reason. Oxide 12 has just had an outstanding track record around the world.

13 over about half a billion tins have been irradiated and the 14 failure rate has been something on the order of .01 percent. ,

15 And those have been very innocuous sorts of failures. So r

i 16 that's an awfully solid fall back position to have.

17 One other point I wanted to make, before leaving the 18 chart, we go with a 950 degree reactor exit temperature. The 19 reason for that is that it allows you to approach fossil plant 20 efficiencies. It's about four to five percent higher in plant 21 effici,ency than one would have if you had a lower reactor exit 22 temperature and a saturated steam cycle. That translates into 23 capital cost savings, substantially less heat transfer surface 24 required, that sort of thing.

25 At the same time, the 950 is low enough that you

20 1 don't get into structural difficulti'es. You don't have 2 attrition of long-term mechanical properties.

3 Can we have the next chart, please?

l 4 [ Slide.]

5 COMMISSIONER BERNTHAL: I give up, what's a-Benson 6 cycle?

7 MR. MCDONALD: That's just a once-through cycle 2

8 where feedwater comes in one end and superheated steam comes

+

9 out the other.

10 COMMISSIONER BERNTHAL: As opposed to seltzer, or

]

i 11 whatever it was?

12 MR. MCDONALD: Well, solser has an intermediate 13 little recirculating loop that still, at the end, comes out l

14 with superheated steam. A saturated cycle is one where the 15 steam that goes to the turbines is saturated and comes off --

! 16 COMMISSIONER BERNTHAL: I also wanted to inquire s

17 about the choice of pool. I'm willing to accept the world's 18 judgment. I've probably seen the words pool and loop 19 juxtaposed over the years about as much as anybody here.

j 20 The argument in favor of the loop, as you know 21 better than I, was serviceability and maintainability like 22 Clinch River was designed that way supposedly. That's very

! 23 important for us, in the safety business. -What's the response

) 24 to that?

4 I 25 MR. MCDONALD: I think, at the time, the Clinch i.

21 i 1 River design approach was settled on', that was certainly the 2 judgment. I think experience since that time has indicated 1

j 3 maybe that wasn't quite right. There's been demonstrated in 4 pool plants overseas very favorable operating and maintenance i

5 capability.

6 Also, in this country, their experimental results 7 have said you're going to be transporting radioactive

'l 8 corrosion products around the loop, distributing them in the 9 piping and various components. And that becomes a heck of a 10 mess to try to maintain. You have to put in temporary 11 shoving, handling that cor.ponents and cutting them out is a 12 big problem.

13 With the pool the main components -- all the 14 radioactive components -- are just sitting in chimneys down in 15 the pool and when the time to maintain them comes, you just I

16 pull them up into an inerted cast, shielded cast. And 17 experience -- I think we've learned a lot and that's why 18 there's a change now.

19 COMMISSIONER BERNTHAL: I thought you didn't have 20 many corrosion products with sodium. You transport corrosion 21 products around?

22 MR. MCDONALD: Fuel planting has corrosion product 23 transport. Some of the projections for Clinch River, for 24 example, which say that -- help me try to remember, Bob. I 25 1 think within a year or so you might talk as high as .3 r per 1 I

22 1 hour out in the IHX area, for example.

2 MR. LANCET: That sounds right. You have a vary 3 large surface area on the fuel pans, and so very small 4 corrosion amounts to a lot of material.

5 MR. MCDONALD: At any rate, when that is going on, 6 within the pool plant itself, it's a much more manageable kind 7 of a thing.

8 This shows just the nuclear island for this standard 9 350 megawatt design. And in developing the design, wa decided 10 on basically four areas. one, this self-protective passive 11 safety feature approach. Secondly, we emphasized trying to 12 e

simplify the configuration and minimize the construction 13 quantities. Third, we wanted to localize and minimize the 14 numbers of systems and components upon which the nuclear 15 safety of the plant depended. And finally, we wanted to 16 facilitate operator activities.

17 And we feel like we've done pretty well in 18 accomplishing that. In a couple of charts, I'll illustrate 19 some of that.

20 Essentially, all of the radioactive material of the 21 system is included in this envelope, this 39 foot diameter, 22 factory fabricated reactor vessel, which has no penetrations.

23 It's a very simple configuration.

24 The exceptions to that are that are that over here 25 we have spent fuel storage. We have this in-vessel storage of

23 1 all the spent fuel for a one year pe~riod that allows 2 radioactive decay to the point where when that gets 3 transferred over to spent fuel storage and then shipping, the 4 power level is sufficient low. It can be cooled by natural 5 draft of air cooling. And that cooling is adequate. In the 6 event that there are any accidents or stoppages or drops 7 anywhere as the spent fuel is being passed through this 8 permanently installed fuel handling system that you see there.

9 This is a fuel handling approach that we adopted 10 from the French Super Phoenix. We've made some improvements, 11 but that's been shown over 12 years now to be vary effective, 12 very efficient.

.i 13 The reactor vessel is centered in a reinforced 14 concrete reactor containment building, which is 57 feet by 57 i

15 feet. The building height is comparatively short, compared 16 with other reactor designs. That's possible because we have 17 the crane outside the building and this crane can serve as 18 other modules that might be aligned with it.

19 Any component handling that's necessary is done 20 through hatches in the roof.

21 You can see the reactor vessels -- or rather the 22 E-exchangers, the pumps that are just hanging suspended within 23 the pool. And the close coupled system steam generators out 24 here, because of this close couplir.g we have -- we use model 25 cleaner piping. And the advantage of that is that it

-- - - . - - ~ - . _. .-- - _ . . . - . ..

24 1 eliminates the need for use of sprin'g hangers and snubbers and i

2 that greatly simplifies the design, reduces the number of 3 potential failures, reduces-the burden for inspection and so 4 forth.

5 MR. MCDONALD: I guess one last point I would make 6 on this chart is that, as we have tried to shove as much of 7 the fabrication, particularly the safety-related items, back f

8 into the factory with this environment of enhanced quality, 9 and also by minimizing the number of parts that are involved 10 in nuclear safety-related activities or functions, we are at a 11 point.now where about 70 percent of all the hardware that has~

12 a nuclear safety-related function is constructed in the shop, 13 whereas in previous designs about 30 percent has been factory

! 14 fabricated.

l 15 Can we move to the next one, please?

16 (Slide]

17 This shows an arrangement in which four of these 18 nuclear islands are aligned to give a 1400-megawatt station 19 capacity, and I wanted to make two points off this chart. By 20 minimizing the size of this nuclear safety-related envelope, 21 we allow ourselves to go to high quality industrial standards 22 for construction of other parts of the plant -- for example, 23 the balance of plant.

24 If one chose to, you could construct these entire 25 sets of separate turbine generator buildings as a separate

25 1 project using industrial standards, completely segregated even O( ,j 2 by a fence, if you like, from the nuclear islands. So that 3 yields a very worthwhile cost savings according to Bechtel's 4 estimates.

5 The second point I wanted to make was that by 6 ganging these together, it is possible to gain some cost 7 savings by sharing facilities, such as the crane that I had 8 referred to earlier, the maintenance facilities, the fuel 9 transporter. This particular arrangement shows a co-located 10 fuel cycle facility for reprocessing and refabrication of 11 metal fuel, and Argonne's estimates for such a facility to 12 support this kind of capacity is that such a building would be 13 about a $53 million cost item.

14 Could we have the next one, please?

15 [ Slide.]

16 I mentioned the emphasis we placed on trying to 17 lower construction quantities, and I think this is an 18 illustration that we have been successful. Here are different 19 construction materials, and I don't want to dwell on the 20 numbers but just indicate that the unit quantities over here 21 -- that is, yards per megawatt, tons per megawatt -- are 22 significantly lower than current generation PWE.s. We 23 certainly expect advance light-water reactors to come down 24 significantly as well. I don't know whether they will do this 25 O well, but a good reflection of the success we have had in

l l

26 1 reducing construction quantities and therefore schedule and  ;

2 costs is that Bechtel now estimates that our craft labor man 3 hours per megawatt for SAFR is getting down pretty close to 4 the cost you would pay for a coal plant, and it's like 30, 40 5 percent lower than you would pay for field labor for a current 6 PWR.

7 Next' chart, please.

8 [ Slide.)

9 Let me make one point before starting to talk about 10 the safety features. We feel that this concept can provide a 11 very worthwhile leg up on other reactor types in the area of 12 availability, and that is because of the undershield refueling 13 system that we have because of the much greater simplicity --

14 fewer engineered safety systems, fewer auxiliary systems, no 15 spring hangers, no snubbers. I have some material in the 16 handout to that effect.

17 We established top-level safety goals for ourselves, 18 which also are in the handout, which actually have demands 19 greater than the NRC and EPRA requirements for dose and risk.

20 If I can have the next chart, please.

21 (Slide.]

22 This cartoon really illustrates essentially all of l 23 the features in the plant that really have a bearing on 24 nuclear safety. There are a few others, but these are the key 25 ones. It is with these features that we are able to meet

27 1 those goals. We supplement the norm'al decay heat removal

) 2 path, the normal heat removal path with two diverse passive 3 decay heat removal systems. One we refer to as RACS. That is 4 the natural convecting air that comes into the reactor cavity, 5 cools the outside of the leak jacket, which is part of 6 containment, and exits.

7 The other we refer to as DRACS. It is direct 8 reactor auxiliary cooling system, where we pull energy out of 9 the pool, natural convecting secondary sodium loop, and then 10 transports the energy over here where it is rejected to air in 11 a natural draft heat exchanger.

12 We have diverse redundant reactor shutdown 13 mechanisms, including one that is a self-actuating system that 14 is being developed on the DOE program.

15 [ Slide.)

16 An important feature I want to address on the next 17 chart are these self-limiting feedbacks, which are the 18 ultimate fallback that will assure the safety of the plant. A 19 number of other features.

20 In general, let me just close by saying that a 21 reflection of the simplicity of this approach, I think, is the 22 fact that the total 1E emergency electrical power required for 23 this system, actually for four of these 350 megawatt units, is 24 about 28 kilowatts, and that compares with about 5000 required 25 in the CRBR design, and about 9000 kilowatts required for

_. . .- . _ .. =---- _-.- - -

28 1 emergency power in a current LWR.

2 Can we go to the next one, please.

3 (Slide.]

4 Any liquid metal reactor can be designed so that it 5 has self-regulating, self power limiting characteristics, and 6 that was illustrated by a test run earlier this year'in EDR 2 7 and tests run a couple of years ago in FrLnce in the Rhapsody 8 reactor. Unfortunately, these acronyms have been switched, 9 but this illustrates the experiment done in EDR 2, loss of 10 flow without heat sink, in which when the reactor was at full 11 power, the pumps wars turned off, so the forced convection-12 eventually stopped. But at the same time, the operator 13 stepped back and simulated the disconnection of the plant

'- 14 control and protection system, so there was no electronic 15 protection being provided and the reactor was just left to 16 take care of-itself, in a sense.

17 ~What happens is that with the reduction in flow, the 18 sodium temperature goes up, and that has two very important 19 effects. One is that it heats up the control rods so that 20 thermal expansion pushes the poison down into the core and 21 that decreases reactivity. Additionally, as the core heats 22 up, it expands and more neutrons leak and that brings down i

23 reactivity. So the power came down, as you see there, and 24 eventually the sodium came down, temperature came down, and 25 finally the power level will equilibrate at a level equal to

.. s at ,nM M 29 1 the heat losses from the system.

2 With the loss of heat sink without. scram, a similar 3 effect, the steam system is essentially cut off and it takes 4 care of itself. The French ran the loss of flow without 5 scram. We have incorporated in the SAFR design the same sorts 6 of characteristics as illustrated by the plots over here, one 7 additional plot for transient overpower without scram, and 8 that is where you pull the control rods out without scramming 9 the reactor, so we have a safe response test.

10 (Next chart, please.]

! 11 I apologize. Let me touch on this and the I will 12 leave this, and I apologize for the schedule problems.

13 This just illustrates that from what we thought were 14 fairly conservative analyses, we can show that the PRA l

15 calculated for the SAFR can meet the NRC goals with a lot of i

16 room to spare, and to close the safety section, I would just l 17 like to say that the SAFR design complies in each and every 18 respect with all the NRC advance reactor policy statements.

19 In the handout, there is a listing of the elements of those 20 statements and describes how the-SAFR design does comply with 21 those requirements.

22 Could we move on to the next one, please?

23 [Sl'ide]

24 Just a summary of economics.

25 Let me.try to quickly get to the question that you l

30 1 raised. First of all, DOE provided the guidelines for a 2 comparison with alternate power sources, and that's where 3 these coal figures come from. This is energy cost as a 4 function of time relative to'the impact of safety on our 5 costs. In fact, by going to this. fairly passive system of 6 safety, avoiding the cost of engineered safety systems, being 7 able to reduce a lot of the instrumentation that is involved, S we have really been able to bring the cost down. So the 9 approach we have taken has worked in'the direction of bringing 10 cost down.

11 We made fairly detailed analyses which were

~

12 independently checked by an outside group based on DOE s 13 guidelines and came up with a conclusion that our first SAFR 14 plant would have a pretty darned good chance of being 15 competitive with coal plant energy costs because of the high 16 cost, fuel cycle cost of coal, of course, or fuel cost of 17 coal.

18 The mature plants, in our estimate, have a pretty 19 good shot at being very competitive, both on a capital and 20 energy cost basis, with light-water reactors.

21 COMMISSIONER ASSELSTINE: To what extent are those 22 estimates dependent upon plutonium recycle?

23 MR. MCDONALD: Back here in the early stages we are 24 counting on a uranium fuel stowaway cycle. They are dependent I

25 to the extent of, I would say, about 4 or 5 mils per kilowatt

31 1 hour, that neighborhood, when you get out to the mature plant.

2 Next one, please.

3 (Slide.]

4 I could give you specific answers if you would like.

5 We are talking about a program which would result in 6 the design and construction of a plant which would be 7 installed and operated on a utility system, but before'the 8 commercial operation would begin, it would undergo a family of 9 safety tests, which would be system-level tests supplementing 10 what ve had done earlier, and this, we believe, on the heels 11 of the some 17,000 megawatt years of experience with liquid 12 metal reactors around the world, would provide a pretty good s 13 basis for requesting the certification of a standard design at

)

14 that point. That, of course, would greatly facilitate this 15 licensing in haarings of subsequent SAFR units because those-16 could focus on site-specific issues.

17 Can we look at the next one, please?

18 [ Slide.]

19 We have a first plant funding approach we worked out 20 with a financial group that we are reviewing with

21. corporate. We think it is very attractive. As soon as the 22 corporate completes its review, we will review this in detail i

23 with DOE and others. The approach we are talking about makes 24 use of the economic value of the plant. That is, you don't 25 throw it away after a safety test. So that the cost to the

32 1 government before beginning operatio~n would be limited to

'h y ,/ 2 $300 million or less. The ultimate cost to the government 3 would wind.up being essentially zero. The annual funding 4 levels would be readily manageable, se believe, a reasonable 5 return to the operating company, and the risks are equitably 6 parceled among the participants.

7 Can we have the last chart, please?

8 (Slide.]

9 We on th9 SAFR team are very enthusiastic about 10 this. We think we have got a real winner. We think it has the 11 potential of alleviating a lot of the problems that exist 12 today. We think the design is certainly capable of providing f-~g 13 risks that are comparable or even lower than light-water

(

\~-} 14 reactors in the same time frame, can be competitive, have a 15- short schedule to be user friendly, and as indicated before, 16 we think although utilities aren't interested in a major 17 financial role now, I think that will come in the future as 18 we get a clearer view of the licensability of the design.

19 So I apologize for running a little long, and that 20 completes the presentation. I will be glad to respond to any 21 questions you have.

22 CHAIRMAN ZECH: Thank you very much.

23 I know we have probably been too optimistic this

-24 morning in the number of presentations that we have planned, 25 but I think we really should move along, and I would ask the f

I

33 1

g- s following presenters to try to be brief. If we can't really )

i

(_,) 2 cover the subject as thoroughly as we want to, Dr. Rossin, we 3 will schedule another briefing. I know my fellow commissioners '

4 will have questions.

5 I suggest we try to proceed and move along as fast 6 as we can. We will schedule another briefing if necessary.

7 MR. ARMIJO: I will try to be brief.

8 Good morning, gentlemen. My name is Sam Armijo. I 9 am General Electric's manager of nuclear systems 10 technology. Also presenting this morning will be Dr. Glenn 11 Sherwood, who is manager of General Electric's safety and 12 licensing operation.

I 13 I would like to talk very briefly'about the main

- 14 concepts that General Electric is using in the design of the

! 15 PRISM reactor. The acronym stands for power reactor 16 inherently safe module.

f .17 Next slide.

l 18 [ Slide]

19 GE has been working on the liquid metal reactor 20 program in support of the Department of Energy for many l

21 years. We have a great deal of experience. Our biggest 22 contribution to this technology was the building, operation 23 and testing of the Southwest Experimental Fast Oxide Reactor

, 24 in the 1960s with the cooperation of the AEC and 25 U.S. Utilities.

l

_ _, ___ , _ . , - - - . - - - - - - - - - - - - - - - - - " ~ - - ' ' ' " - ~ - - '- ~ ' - ~ ~ - - - ' ' - -

34 1 In that 20 megawatt reactor which we built and 2 operated, we ran what we called a safety test to demonstrate 3 the inherent safety of oxide fuel, specifically the doppler 1 4 effect. It was a success ful test. It answered once and for i l

5 all licensing and safety issues regarding oxide fuel, and we 6 think that is a good model for future advance liquid metal 7 reactors.

8 In the early eighties we started a privately funded 9 program to see if we could come up with alternative ways of 10 reducing the cost of construction and operation of advanced 11 liquid metal reactors. The concept turner out to be one that 12 was a modular design, which has many advantages which s

\

13 Mr. Mcdonald talked about and I won't repeat here.

'-- 14 In early 1984 the Department of Energy awarded two 15 contracts, one to General Electric for the conceptual design 16 of an advanced liquid metal reactor that would have a number 17 of economic advantages, but specifically address the issue of 18 enhanced safety. In the liquid metal reactor we believe that 19 enhanced safety can have major benefits with respect to the f

20 economics of construction and are not at cross purposes with 21 .one another if you do it right.

22 We are a little more than midway through the 23 conceptual design. The work is going well. We have the 24 cooperation of the national laboratories as well as other 25 industrial concerns in this development.

. l l

35 1 When the work is completed, we are hoping that the 2 program will continue to the point of what we call a safety 3 test. The safety test would build and test a full-scale 4 prototype module to demonstrate the safety of the machine, the 5 self-pretection characteristics for beyond design basis 6 accident events. We believe that is necessary not only to 7 convince the professionals in the industry but, more 8 importantly, to convince the general public that we truly have 9 a safe reactor.

10 MR. ARMIJO: Next chart.

11 [ Slide.]

i-12 MR. ARMIJO: Our design approach is a compact pool

13 reactor. We have always believed that the pool design has

}

14 advantages. It's been a long debate. You've heard all those 15 arguments, pros and con. But the pool reactor gives us the 16 advantages we're looking for, and that is containment of the 17 safety systems within the reactor module itself and not 18 distributing the radioactivity throughout the system.

19 The issues of factory fabrication have already been 20 covered, the advantages. Our design is smaller than the 21 design talked about previously, and we believe that has 22 several advantages with regard to economics and safety. One 23 of the most important advantages, of course, is if you have an 24 economical plant that is small, the cost of demonstrating it 25 is also lower than the larger plant.

l l

36 1 Like all modular plants whi'ch will be discussed 2 today, they provide the utility.with a lot of flexibility with 3 regard to adding capacity as he needs it, rather than making a 4 multi-million dollar commitment for a need that's 10 or 12 5 years down the road. And of course, I mentioned the I 6 safety-related equipment being limited to the module.

7 Next chart.

8 [ Slide.]

l 9 MR. ARMIJO: I'm not going to go into a lot of

.i l 10 detail. But just to familiarize you with what the module 11 looks like, in schematic form. The reactor is about 60 feet 12 high, about 20 feet in diameter. All the safety-related i

! 13 mechanical equipment is in the pool. The core is at the 14 bottom submerged under many feet of sodium. It has two 15 intermediate heat exchangers that transfer the heat from the 16 radioactive to a non-radioactive sodium system, and it has 17 four electromagnetic pumps that pump the sodium without any 18 moving parts.

19 It has a thick wall for this low pressure system,

! 20 one to two-inch. thick stainless steel reactor vessel. And it 21 has an outer containment vessel also of the same thickness.

i 22 The top head is hermetically sealed, all welded seals up in 23 the top heads and there are no active systems taking 4

I

! 24 radioactive materials in and out of this system.

25 7 I'll describe later another feature of this machine R

i

37 i

I 1 is that the reactor is continuously' cooled. The containment 2 vessel is continuously cooled by a flow of air coming down 3 from the outside and around the reactor vessel. During normal 4 operation it takes out a little bit of heat. When the 5 temperature of normal operation, which is well -- the heat 6 loss is very small.

, 7 But in the event you have a thermal transient, the 8 temperature of the reactor increases, the efficiency of 9 cooling by this radiant vessel, auxiliary cooling system rises 10 as temperature to the fourth power. A very effective passive 11 means of cooling the containment in the event of a thermal 12 transient.

13 Next slide.

O, 14 (Slide.]

15 MR..ARMIJO: Our concept is to design the plant, the 16 modules so that you can contain the safety-related equipment 17 within this boundary. So if you just look at that, we have i

18 three modules feeding three steam generators feeding one i

j 19 turbine generator system. This is about a 415 megawatt 1

j 20 electric power block. Building incrementally, then you.can 21 add independent power blocks the size you want.

22 The modules themselves contain all the r

23 safety-related equipment necessary to shut the plant down and 24 to protect the plant. The balance of plant, which you can see 25 is about two-thirds of the area of this plant is i

i

, _ . ~ _ . . . . ~ . . . . - _ - _ , . _ .- . _ - , _ - - - , , .- -. _

38 1 nonsafety-grade because its performa'nce and operation is not f 2 important to the safety of the system. And our concept is .

1

'3 that if we can prove that by~a safety test demonstration, we '

l 4 can obtain the economics that we get from non-nuclear grade  !

5 construction in the balance of plant.

6 Next slide.

7 [ Slide.]

8 MR. ARMIJO: Just to give you an idea of another 9 feature of this reactor is this elevation view. As you can 10 see, the reactor module is below grade in a silo,' as is the 11 steam generator and sodium water reaction protection system.

12 There is also a large separation between these two systems of 13 over 100 feet. The reactor module, because it is small, is

\

also seismically isolated, and it takes away the major 14 15 contribution to seismic challenges to the system. And that's i

16 built into our design.

17 Next chart.

18 [ Slide.]

19 MR. ARMIJO: Now we have a conventional design basis 20 requirements. We meet all the NRC top level regulations with 21 regard to the design of such a plant. We provide equipment 22 for highly reliable safety-related shutdown. We have a scram 23 system that's got a high reliability built into it. We have 24 shutdown heat removal system, which is this passive system 25 which I described. And we have containment. These are pretty

39 1 much the traditional approach.

('_)\

N,, 2 Beyond that -- next chart.

3 [ Slide.]

4 MR. ARMIJO: In the beyond design basis event, what 5 would normally be called the ATWS event, we handle that with 6 passive -- with a passive response. We believe that the 7' reactor shutdown, the reactor cooling, decay heat removal and 8 containment all can be performed passively with this reactor 9 module.

10 And we've analyzed it for several events -- I'll 11 describe three. First, in the case of loss of primary or 12 intermediate or steam water cooling of the system, the reactor

-s 13 will shut itself down even if the active safety systems don't

\'- 14 operate. And this is a phenomena that I'll describe in just a 15 little bit of detail later.

16 In addition, in the case of normal heat sink without 17 scram the reactor will shut itself down and cool itself 18 passively. Now if you'd back to that back-up chart.

19 [ Slide.]

20 MR. ARMIJO: First of all, what happens if you lose 21 the heat sink, the cooling from the secondary system. What 22 happens is the reactor coolant starts to heat up, the sodium 23 continues to flow in natural circulation. And that'will 24 happen whether the pumps -- even if the pumps are off. So if 25 you've lost all off-site power, sodium would continue to move l

40

~

1 with the natural circulation.

2 The temperature goes up, the reactor core starts 3 thermally expanding because of heat. The bottom plate starts 4 thermally expanding. The control rods thermally expand and 5 the reactor shuts itself down in a few seconds. Not as fast 6 as a scram system, but in a few seconds.

7 The containment is now radiating heat very 8 efficiently, and that containment then takes away the heat and 9 you're at a safe and stable condition. So this is the basic 10 concept of this core being cooled by natural circulation of 11 sodium, and natural circulation of air without any off-site 12 power and without any active systems needing to operate.

13 Next chart.

14 [ Slide.]

15 COMMISSIONER BERNTRAL: You say it's seismically 16 isolated, and that's a3 fairly simple engineering task because l 17 of the silo arrangement?

18 MR. ARMIJO: Yes. The purpose of the seismic 19 isolation is twofold: to reduce the amount of seismic 20 constraints that you have to put inside the structure.

21 COMMISSIONER BERNTHAL: Sure.

22 MR. ARMIJO: And also to make the reactor amenable 23 to siting at various locations with a minimum of redesign.

24 COMMISSIONER BERNTHAL: Sure.

25 MR. ARMIJO: This schematic shows you what our

41 1 -passive air cooling looks like. It's important that you see 2 this. We have a capability for bringing large quantities of 3 air through very large openings which are always open and 4 there's never any delay.

5 So that this system is very efficient in reducing

6. the temperature of the containment no matter what kind of 7 thermal transient we put into it. Blockage of this system 8 would have to be over 95 blockage before we start losing

'9 significant efficiency here. It's a very efficient system.

10 Back to the charts, 11 [ Slide.)

12 COMMISSIONER BERNTHAL: Either of these two cooling 13 mechanisms is sufficient in itself then essentially? I don't 14 know if I want to say either.

15 (Commissioner Roberts left the room.]

16 COMMISSIONER BERNTHAL: If you lost in seismic

}

17 event, if you lost the natural convection around the jacket of t

18 the steam supply system, the natural convection of the sodium 19 in itself would be sufficient then?

20 MR. ARMIJO: The natural convection of the sodium is 21 important to get the heat out to the containment vessel. Then i

, 22 you need the air convection on the outside to get the heat out

, 23 completely.

1 24' COMMISSIONER BERNTHAL: But in that sense you're not 25 -- you're cooling system is not truly seismically isolated.

42 1 MR. ARMIJO: No, the wholeTreactor is seismically m

2 isolated. The whole cooling system within the reactor --

3 COMMISSIONER BERNTHAL: No, I'm talking about the 4 air flow, the path around the outside.

5 )G?. . ARMIJO: Yes, that part'is not seismically 6 isolated. That's a very large stout building.

7 Now I listcd just the key areas -- or not the key 8 areas -- the more important areas from our discussion today on 9 how this reactor design responds to the policy statement for 10 advanced reactors. And as far as' reliability and complexity, 11 we believe this passive natural air cooling system, which is 12 testable full scale, is a very -- fully compatible with that.

13 We've reduced enough complexity because we've 14 reduced the number of shutdown systems required. For example, 15 in the PRISM reactor, we have one set of primary scram rods.

16 We don't have a secondary scram system, although there is 17 redundancy and diversity within that one set.

18 Our PRA assessments to date indicate that we meet

! 19 and comply with all the design goals that NRC has stipulated.

20

[ Commissioner Roberts returned to the room.]

21 MR. ARMIJO: Margin, very important factor on i

22 enhanced margin. We have deliberately kept a low temperature l

23 with our sodium system of 875 degrees F. That's over 100 I

24 degrees lower than what was chosen for Clinch River. And we 25 run with a saturated steam cycle.

l l

43 1 We believe that the added cost and complexity of a 2 higher temperature, sodium temperature and steam temperature 3 actually is not an efficiency. When you add everything up, 4 you wind up paying for it in safety systems further down the-5 stream.

6 We use a metal fuel core because it has advantages 7 of inherent safety, and.we believe will have advantages on the 8 reprocessing of the fuel. And of course, I mentioned seismic 9 isolation.

10 MR. BROWN: Let me add, that the answer to the 11 seismic isolation was not correct. The RVACS concrete stacks 12 are on the seismic isolation. So all of that heat removal N 13 system is seismically isolated.

14 COMMISSIONER BERNTHAL: So the entire convection, 15 air convection --

16 MR. ARMIJO: The stacks.

17 MR. BROWN: The whole air passages are all on 18 seismic isolation.

19 COMMISSIONER BERNTHAL: It seemed like a 20 -vulnerability otherwise.

21 MR. ARMIJO: The outer walls --

22 MR. BROWN: The outer walls are not, the furthest 23 extreme outer wall. But the whole air passage --

24 MR. ARM JO: Good point. Thanks, Neil. Next chart.

25 [ Slide.]

44 1 COMMISSIONER ASSELSTINE: You mentioned recycle.

2 Are you depending upon recycle for the economic viability 3 of the option?

4 MR. ARMIJO: Yes, we are. We do not believe that a 5 liquid metal reactor fueled strictly with uranium as a 6 throwaway cycle would be economical.-

7 CHAIRMAN ZECH: Do you have a positive or a negative 8 temperature coefficient?

9 MR. ARMIJO: We have a negative temperature 10 coefficient. This reactor shuts itself down as the 11 temperature and coolant increases.

12 CHAIRMAN ZECH: All right.

1 13 MR. ARMIJO: We use simplified, with respect to 14 passive systems. Our radiant vessel auxiliary cooling system, 15 which I've described, is a passive system. It's always 16 operating.

17 We have a sealed and inerted containment system

, 18 which has no active penetrations. The simplicity and 19 complexity is being removed systematically.

20 And with regard to severe accidents, we have highly

21 reliable shutdown systems. But most important, we can 22 demonstrate full scale that this reactor can protect itself in 23 a safety test, and that's more important than whatever we can 24 claim.

I

45 1 [ Slide.]

2 MR. ARMIJO: Now regarding that safety test, I can 3 tell you that we're right now between the contractors and the 4 national labs in the R&D phase. Particularly in the metal 5 fuel area, Argonne National Laboratory is working to 6 demonstrate the performance and reprocessing capabilities of 7 the fuel.

8 To. bring that all together, we need a focus. We 9 believe a safety test is the way to.do it, in which the fuel 10 reprocessing skills and experience of the National Laboratory 11 would fuel the reactor and refuel the reactor once it was 12 operable.

13 On completion of the safety test, we would go into a 14 power demonstration phase and our concept is, at this point, 15 we move from essentially all government funding to essentially 16 all private sector funding, if we now have a power plant, if 17 the safety questions have all been addressed and answers that 18 effectively.

19 Now I'd like to turn it over to Glenn Sherwood, to 1

20 talk about our licensing program.

21 COMMISSIONER BERNTHAL: If I may ask one question, 22 both of these liquid metal cooled systems have a positive void

23 coefficient. As I recall, the Clinch River core was 24 redesigned at one point for a variety of reasons, among which 25 l was the elimination of the positive void coefficient, I

i 46 1 believe.

2 Granted, you have a number of inherent safety 3 characteristics here that that plant could not boast, but what 4 is it that makes you comfortable with that positive void 5 ' coefficient, especially in light of the recent publicity that 6 . hat kind of characteristic has received?

7 MR. ARMIJO: You'd have to get into sodium boiling 8 phenomena, which as I mentioned earlier, the sodium boiling 9 temperatures are up around 1600 or 1700 degrees.

10 COMMISSIONER BERNTHAL: I understand.

11 MR..ARMIJO: The peak temperatures that we see, even 12 under these very unlikely events, are much less than that, in

-~g 13 the 1100, 1200 degrees Fahrenheit, even without scram.

'- 14 COMMISSIONER BERNTHAL: So you simply don't consider 15 it a credible accident, that the voids would appear in the 16 core?

17 MR. ARMIJO: The only mechanism to create voids, 18 that we know of, would be boiling of the sodium.

19 CHAIRMAN ZECH: That's why you call it a negative 20 void --

21 MR. ARMIJO: No, the question I answered to you was 22 --

23 COMMISSIONER ASSELSTINE: It's a positive void 24 coefficient.

25 MR. ARMIJO: For voids. You asked about

47 1 temperature, I believe. '

~

2 CHAIRMAN ZECH: All right. But the void is 3 positive?

i 4 MR. ARMIJO: Correct.

5 COMMISSIONER ASSELSTINE:. Right.'

6 COMMISSIONER BERNTHAL: What about a loss of 7 coolant?

i 8 COMMISSIONER ASSELSTINE: Or flow blockage like 9 occurred at Fermi 17 10 MR. ARMIJO: In the case of a flow blockage there 11 might be some local damage, but that would be a local event.

12 But that wouldn't cause the whole reactor to do anything 13 significant.

'14 MR. SHERWOOD: Can I have the first chart?

15 [ Slide.]

16 MR. SHERWOOD: What I'd like to talk about is the 17 way that we plan to license this type design with the NRC and 18 with DOE with the cooperation of the laboratories. Dr. Armijo 19 mentioned the inherent safety capability of the PRISM system 20 and this will essentially form the entire backdrop for the 21 licensing process.

22 The PRISM design obviates most of the regulatory 23 requirements for light water reactors as we've known them in 24 the past. This design has no diesel generators, it has no

25 miles of safety-related piping and pumping of high pressure or

- , - , . - , _ _ _ , ,_ - - , - . - , - -------r-- -- - - - - - - - - - - - - - - - r -- - < - - ' - - - --+--r-

48 1 low pressure, it has no safety-relat'ed control room.

2 Essentially, it's limited only.to the PRISM module, as 3 Dr. Armij o said.

4 Therefore, we feel that the way to test this -- the 5 way to. license this is by test, somewhat analogous to the way 6 that we might even license the GE CF-6 engine. In other 7 words, although we won't kill chickens in this, this would be 8

essentially totally tested as I'll describe in the next couple 9 of charts. It will even be reusable after essentially a 10 fairly severe range of accident-type testing.

11 So it's a unique concept and a unique opportunity 12 for us to go beyond the current licensing basis, essentially 13 by analysis and by paper, to go to a-physical test.

O- 14 Essentially the safety test, as Dr. Armijo said, is 15 based upon the inherent safuty features. I think he's 16 probably covered enough of this. I think the bottom line is 17 that the system tends in all accident scenarios to 18 equilibrate, to self-equilibrate with the air cooling system.

19 And we see of no way, other than possibly an earthquake which 20 might be five times the SSE, of doing any real significant 21 damage.

22 We feel that also the certification process is 23 ready-made for this.

Given the test and all of the other 24 licensing that the certification will then give the public the fg 25 confidence that we have a truly safe design, and one which can

i 49

'l be made available for commercializat' ion.

s 2 Next chart.

3 [ Slide.]

4 From the point of view of the test activities,'what 5 we would do with the safety test is essentially go through the 6 entire range of the test which one might expect from a light 7 water reactor, in addition to the severe accident test, and in 8 the severe accident test, we would essentially do tests with 9 scram, and these would be typical loss of flow, loss of heat 10 sink, rod withdrawal, as well as the Atlas type tests. All of i

11 these tests, since the corridor span shuts down extrinsically, 12 as Dr. Armijo said, we would be able to go through all of 13 these accident related scenarios and with inspection after the 14 fhet, presumably bring this unit back to commercial operation.

15 It is unique in the sense that it does not self 16 destruct during an accident type scenario. We also feel the 17 tests will demonstrate the factory capability for fabrication 18 and assembly and a modular construction approach for reducing 19 costs.

20 (Slide.]

21 MR. SHERWOOD: I wanted to spend a couple of minutes 22 telling you something about the unique plant protection 23 system. The plant protection system is shown in these two 24 concentric spheres, the outer one being the PCS or plant 25 control system. The other one being the reactor protection

50 1 system, which you are all familiar with, with light water  ;

g j 2 reactors. j 3 Essentially, the reactor protection system is safety 4 graded and limited to the module only. It is totally 5 autonomous. It cannot be changed, controlled or impeded in 6 any way by the operators. The only thing the operators can do 7 is trip it. In other words, to shut down the reactor.

8 Outside of that, there is a plant protection system, 9 which is the traditional control room protection system, which 10 operater tne plant, fits above the plant protection system.

11 This is commercial grade. This is also totally automatic.

12 The plant can start up on computers, can operate or computers

-, 13 and will shut down on computers.

14 There is a number of diagnostics that are usually 15 and traditionally provided to the operator in the control 16 room, which she can use to control a plant, but doesn't need 17 to because it would be controlled automatically.

18 There are two separate ways to control a plant. One.

19 is'from the traditional control room and the other is the 20 inherent plant protection system, both of which are totally 21 automatic.

-22 This system is essentially a four channel multiplex 23 with essentially a high degree of redundancy and also a self 24 test. In the normal operation, there are four channels, three 25 on and one idles, and the one that idles circulates among the s

51 1 four. It is an extremely safe although complex system.

2 [ Slide.]

~

3 MR. SHERWOOD: Because of this new licensing 4 opportunity and the fact that most of the rules and 5 regulations really do not apply to the present test, we 6

discussed with NRR the notion of a new rule which we termed 7 Appendi:: S. This would be similar to Appendix 0 in that it 8 would permit applicants to reference the typical prototypical 9 safety test for licensing their plant. Since this is a total-10 plant design and a full test would be of final design concept 11 only. It is consistent with all of the current policies on 12 designs, future designs. I will discuss in a subsequent chart 13 how we will fold this into the schedule.

O N/ 14 [ Slide.]

15 MR. SHERWOOD: One of the advantages of the Appendix 16 S is it reduces the uncertainty with regard to the applicant.

17 The Appendix 0, while it addresses itself to the ref.

18 really leaves -- it is fairly skinny in terms of describing 19 the total concept of how licensing would be done, using the 20 prototypical test approach. This new Appendix S would fix 21 that by describing the process in detail. It would al'to 22 encourage the development of all reactors, not only the GE 23 PRISM, but any other reactors which can be suitably licensed 24 through the testing process could use these. There is no 25 ownership necessarily to General Electric.

~

$2 1 This specifically identifies the role of tests as 2 opposed to paper' analysis. We feel it establishes an 3 efficient alternative for doing the job for advanced reactors.

4 [ Slide.]

5 MR. SHERWOOD: This is the schedule for licensing of 6 the PRISM plant. I will also show you how Appendix S would 7 work. All of you are familiar with a PSAR and the FSAR 8 system. This would me divided into essentially three 9 aspects. One is the submittal of the preliminary safety 10 information document. This is sort of a PSAR. Maybe another 11 way to describe it'is it is only a foot thick, but essentially 12 it's thick enough to describe really what the plant is all 13 about and it covers a preliminary safety analysis and also 14 preliminary PRA.

15 This has been in development for a number of years.

16 We have had a number of meetings with the staff on several 17 ~ concepts. This_will be submitted within the next month or so.

18 As a result of the initial review of this PSID by 19 the staff, since it is a totally new design and concept, they 20 will write some. policy papers for the commission's 21 consideration. At this stage, they are considering one on 22 standardization, containment, accidents, and then they will

{ 23 encapsulate all that in a safety evaluation report.

24 Presumably these would deal with the acceptability of the 25 containment and our envelope that we propose, which is safety

, . . =_ -._- . .___

53 1 grade, which essentially is limited 'only to the module and the 2 re-fueling equipment.

> 3 [ Slide.]

4 MR. SHERWOOD: This shows the schedule for the 5 safety test being designed now. It would be finished at this 6 ' stage of the game, submitted. The test itself would be 7 roughly a two year time period as shown, in the 1995 to 1997 8 time period.

9 The proposed rule, Appendix S, has already been 10 submitted to NRR. They have it under advisement. We would 11 hope it is approved by the and of 1988.

12 The other part of Appendix S calls for a design i

i 13 report. You might think of a design report as something N- 14 equivalent to the FSAR.

This design. report would take 15 everything that comes out of the initial PSID review, take the 16 information we have gotten out of the design and safety test 17 and then put it together into this design report, which would 18 be a detailed report, and maybe this would'be five feet tall.

19 As a result of that, we would then have our safety 20 evaluation report, then our test report, and finally a 21 certification around 1997 or 1998.

22 (Slide.]

i 23 MR. SHERWOOD: Appendix S, as we have discussed it

'24 with NRR, encapsulates this total process. We feel this is a i

j 25 sound licensing program for a PRISM being a design which

54 1 intrinsically is fail safe, intrinsically shuts itself down, 2 doesn't need the diesel generations and all the other

! 3 trappings that we have in LWR's.

4 We have had a number of NRR interactions in the last

.r 5 two years and our technical meetings are underway. As I said, 6 we have submitted the Appendix S.

7 I put this'on there a month ago, but I don't really 8 think this is an issue. I think this has been resolved 9 between DOE and the NRC. We feel that the PRISM concept 4

10 provides an unique opportunity for the industry, for the NRC 11 and for the United States to have a quantum jump in safe'y t and 12 efficiency and cost, and the safety test is the way to do i

13 that, to put a definitive cap on all the questions.

14 Thank you.

15 CHAIRMAN ZECH: Thank you very much. We will move 16 into the next. I believe this is the last presentation, 17 Dr. Rossin?

18 DR. ROSSIN: That's right. Then we will stand ready' 19 for questions until you adjourn.

20 CHAIRMAN ZECH: Thank you. Proceed, please. I must 21 leave at 11:45 p.m. If we can't complete it, I will ask l 22 Commissioner Roberts to take over for me and try to wind it l 23 up. Proceed, please.

24 MR. NORTHUP: Mr. Chairman and Commissioners, I am 25

[' Gene Northup, General Manager of GA Technologies, Inc.'s power l

55 1 reactor program, specifically responsible for the modular HTGR 2 development for advanced reactors.

3 It is my pleasure to be before you today to let you 4 in on how our program has been going and how it has been

-5 rapidly accelerating. We.think we know where we are going.

6 We have hopped down requirements being specified and 7 recognizing that, we will hold to our 20 minute presentation.

8 We are developing a second generation reactor that 9 we think addresses and we believe addresses the public needs, 10 the Government's needs, the utilities' needs, as well as our 11 own vendors' needs for investment protection, as well as 12 something that will meet the community at large.

13 This design that we have been working on is reaching 14 a point where we'can be very proud of it. I hope when I 15 finish today that you will see how far we have come.

16 I'm going to use your advanced policy statement as a 17 way of reflecting the progress on the work. If I could have 18 the first chart.

19 [ Slide.]

20 MR. NORTHUP: Look at the policy objectives that are 21 shown here on the left, it calls for enhanced margins of 22 safety. The modular HTGR not only meets enhanced margins but 23 it meets PAG limits at the exclusionary or boundary with no 24 off-site sheltering or evacuation required. This means the

(N 25 plant boundary of 1,500 feet, there is no concern for

\

! l 56 '

1 evacuation or sheltering.

2 We are using passive safety features to accomplish 3 this. I will explain those a little further later.

I 4 The policy asks for NRC' interaction,-and we have had 5 12 NRR meetings to date. We have had two ACRS meetings and we

'6 have many more scheduled in the near future with NRR. We have 7 had'eight major submittals, including a licensing plan, which 8 the NRC has commented on. We have a top level criteria before 9 them. On September 30th, we submitted a four volume PSID i 10 document for their review on the safety analysis of the plant 11 and description of the plant.

12 We have a future submittal of a PRA, which will be 13 in within the next two months, which fully outlines the~

14 probabilistic risk assessment of the plant and a technology j 15 plan which defines the tests and investigation requirements 16 that are required to support the design.

j 17 I might note we are after a preliminary safety 18 evaluation report and a licensibility statement sometime early 19 in fiscal 1988 or sooner, if we can got it.

20 [ Slide.]

21 MR. NORTHUP: The policy asks for less complex

{ 22 shutdewn and decay heat removal. The modular HTGR has a 23 shutdown ~ capacity of a very large negative temperature 24 coefficient, and this is probably about double any other s 25 competing system. There are two diverse gravity insertable

57 1 control systems, shutdown systems. .There are two decay heat -

2 removal systems which are active, one is the main loop for

3 normal cooling or normal plant operation. There is a shutdown i

4 loop which is there for higher plant availability. That is 5 for shutdowns to get a quick turn around, and there is a 6 separate passive natural circulation heat removal system, very 7 similar to the one that was explained on the PRISM concept, 8 for cooling the modular HTGR. It also is located in c silo.

9 The policy asks for longer time constants for 10 accident management, something that gives an operator not 11 perspiration but time to think about what he is going to do.

12 We have many days before action needs to be taken, if any 13 action needs to be taken at all. This comes about because of 14 the high heat capacity of the reactor core, the los power 15 densit'y that is used for the core, and even when there is an 16 event, the peak fuel temperatures, they peak well below any I

j 17 damaged level to the fuel. We have over 600 degrees i

18 centigrade r7rgin between~what our peak temperature appears to 19 be and any accident versus what the capability of the fuel is.

20 [ Slide.)

21 MR. NORTHUP: Simplified safety systems to reduce 22 operator actions. There are no operator actions required for l

l 23 shutdown and heat removal. The control-room, we believe, is a 24 non-classified or non-safety class system. The operator is  ;

25 there to watch what is going on, not to control an accident.

i

)

l I

r_. . _ , _. , _ _ . . _ - . ,.m _ . . _ _ _ , _ _ _ _ _ . . , , _ . , , . _ . _ _

58 1 Minimum for potential for s'evere accidents, as I 2 mentioned, with'a very strong negative temperature

)

3 coefficient, it controls itself. The loss cf coolant is 4 totally acceptable. We do not require the coolant to even be 5 there to obtain the cooling. Therefore, we do not have to 6 protect against loss of coolant. This coolant is totally 7 inactive with the neutron radiation. It does not interact 8 with~the core. It does not interact with the metals of the 9 system. There is no phase changes or no boiling that occurs 10 with the coolant.

11 The policy' asks for a reliable BOP or safety systems 12 that are independent. The modular system also shows 13 separation of a nuclear island from the balance of. plants.

14 The nuclear part has been contracted to a very small area of 15 the plant, leaving the BOP to more conventional construction 16 techniques and also operation isolation from the rest of the 17 plant. We can have transients in the BOP that have no impact 18 on the nuclear system.

19 Low maintenance and personnel exposures; the high 20 quality fuel that the HTGR's have always been after, certainly 21 retains the fission products. We have low helium and primary 22 circulating activity. It is easy for accessability. We have 23 cases of experience where we have taken equipment directly out 24 of the core area of an HTGR, circulators, for example, control 25 rod drives, and worked on them by hand, because they are easy

59 1 to maintain and not contaminated.

2 Operator exposure, maintenance exposure in plants, 3 from experience on operating plants, it is at least 100 if not

4 more or less than any operating plant today.

5 The next slide a policy calling for defense in depth 6 with multiple barriers and reduced accident potential and a

7 consequence. And here, although I know you're familiar with 8 the fuel and the system I'm talking about, perhaps for the 9 audience I could paint a picture.

10 If you take a secondary containment structure that 11 houses a regular light water reactor and shrink that down 12 around a small particle of uranium, and then you take three 13 more structures just like it and shrink those down around the 14 previous one, you have a multi-layered, shrunken vessel 15 housing that uranium kernel. '

16 If you'll now picture that that capacity of that 17 structure is equal to 50 or more times the capacity of an 18 existing secondary containment, you may get a mental picture 19 of how significant the HTGR fuel is. We then devised the 20 system surrounding that fuel such that it never causes those 21 structures to be subject to failure.

22 In addition, this shrunken vessel is put into I

23 graphite so it's directly in contact with this moderator. It 24 is inside a steel vessel, it is inside a silo enclosure below 25 ground, so the items of defense in depth are very strongly 4

,p-,4 ~

__ _ . . - ,me_

I 60 1 included within the modular HTGR design.

( 2 Prevention and mitigation. We find that it's highly 3

reliable in.its operation, and it's operator-friendly, a very 4 forgiving design. Negligible releases are proven by a proven 5 technology. And for proven technology we have the Magnox 6 reactors which have operated for many, many years. The AGR, 7

the AER which is'the German test reactor in German that's been 8 running for 19 years. This reactor was run through a loss of 9 coolant test 15 years ago and proved the fact that the 10 gas-cooled reactor could be cooled by natural convection 11 cooling.

12 The THTR which just came to 100 percent power 13 operation in Germany last month also is providing an 14 experience base. Peach Bottom 1, Fort St. Vrain, all of these 4

! 15 reactors have provided both bad and good E..eperience, and the i

16 . experience with them have been put into the modular design.

17 On the fuel particle, the reason we believe it is so l

18 good is that we've had actually billions of particles that

) 19 have been run in reactors and have been examined and tested.

! 20 In addition, this last' year we finished 50 KG of high quality 21 fuel equal to anything that the Germans have produced.

It was 22 purported that the Germans were the ones who could produce 23 high quality fuel. We produced 50 KG last year in a pilot 24 line, full segment scale line of GA, and this fuel fully meets 25 the specification requirements for the modular reactor.

s l

nw-- - - - - - -- - - - , - - - ,,m---g --p,,,, w .g-,.,,,-..,m,,.-,,,_yy, , , ap-- -,m,--,-ae,-.---- -,-,c.y,c,.--. men,--ww-.,v-,,---- ,,,.m--- ---,--.,----._----,---m--,,e

61 1 It's a low-enriched throw-away fuel cycle, it's 19.8

\ 2 percent enriched uranium mixed with thorium. We never have to 3 exceed the enrichment. We fully meet the proliferation 4 requirements, and reprocessing is not an issue to gain our 5 economics. ,

6 Now, the --

7 COMMISSIONER BERNTHAL: Can you go lower than that

! 8 on the enrichment without paying a significant price?

9 MR. NORTHUP: We could go to lower enrichments and 10 cut back on the use of thorium and even maintain the same 11 condition. But we find that using it at a maximum with 12 thorium included allows us to balance the mixture throughout I

s 13 the core, rather than having different enrichments for each

' N- 34 segment'of the core.

15 COMMISSIONER BERNTHAL: I'm asking, as you know, 16 because 19.8 is a number that's clearly chosen advisedly and l 17 it's not clear that real life is quite so sharp.

18 MR. NORTHUP: No, that can be varied.

19 Two things that are not on the advanced reactor 20 policy, of course, that are very important to us are cost and 21 schedule. With regard to coct, we're finding that the cost 22 projections of the modular system are 20 percent or more below 23 a coal plant of equal size. On the might water existing, l

24 certainly current light water picture, the best light water

(h 25 picture at least 10 percent lower in capital cost, in m -- - - - , - - - , - - , , - - - , , - - - - , ,--,-.,,.-,-p_ ,, . - - - - ---------------,e-,-y, -

-mm -

-_ _ . . - _ - . . _ _ . .. - . _ = .

62 i

1 operating cost, power cost.  !

l

! 2 The other aspect of build quickly, we've 3 incorporated all the aspects of shop fabrication and quick i

4 site installation, and Bechtel is projecting on equilibrium 5 plants from the start of site construction to plant operation 6 of 27 months.

7 The modular ACGR's program we think is well on the 8 track. Dan Mears is here with me today -- he's Chairman -- to I

l 9 talk about the utility perspective. I would add before he

-10 starts on that that the utility has been -- the utility group,

+

11 Gas cooled Reactors Associates, has been very supportive of 12 this program. They represent about 30 percent of the 13 generating capacity in the United States, and they have been 4

i 14 working with us on criteria development, talk-down criteria, 15 checking of the design, and have been very supportive.

16 GA Technologies, my company, was just recently 17 purchased by a private family and they have taken the HTGR on l'

i 18 as a major item to pursue. They're giving it their energy and l 19 attention.- Of. course, they, too, recognize that in areas of 20 high risk, as this type of program is, they need the help of 21 the government and the utilities working to that end, but

(

t i

22 they're fully behind the program.

l 23 CHAIRMAN ZECH: Thank you.

I

24 MR. MEARS

Good morning. My name is Dan Mears and 25 I'm General Manager of Gas Cooled Reactor Associates and I'm i .

33 g 1, pleased to be here with a few members of our staff and several 2 members from the utility industry that are on our licensing 3 development subcommittee. As Gene said and as you well know, 4 Gas Cooled Reactor Associates I believe is the remaining, lone 5 standing utility user group in support of a specific advanced 6 reactor concept.

7 We have been in existence since 1978 and have played 8 a key role in the development of utility user requirements, 9 assessing the design to make sure they respond to those 10 requirements; providing a coordination, integration, support 11 role with Department of Energy; and of recent, have been very 12 active in the development of a project initiative for the 13 modular gas-cooled reactor.

14 Going back in the early formative days of this 15 concept, --

16 (Slide.]

17 we performed a bit of a grass roots effort within the utility 18 industry to survey the direct utility input for key features 19 and issues associated with all the advanced reactor concepts.

i 20 This chart is just representative but indicative of the l

21 particular input that we've received in our efforts to bring 22 these particular utility user requirements to bear. This one 23 addresses the size of the nuclear plant that's preferred by j 24 the utility industry, and indeed, does confirm that the size 25 range that we're looking at here with the advanced reactor

64 1 concept is on target with the input that the utilities are 0,

1 2 prepared to back up.

j 3 (Slide.]

, 4 This is just a summary of some of the key specific

.o 5 utility requirements that have evolved now over the last 6 couple of years that we have been in interaction with the

.\

) 7 design team on this program trying to find an optimum, of 8 course, with regard to safety, reliability and economics.

9 I don't have time to go through each point but I 10 would note that we were the first group to acknowledge the 11 ability and to make the requirement to preclude the need for 12 sheltering and evacuation plans with this concept, which is 13 being maintained through the design and analysis process. At 14 the same time, maintaining an economic plant which, of course, 15 we're sensitive to and have been making these comparisons on a 16 continuing basis with advanced coal technology.

17 (Slide.]

18 Within the last 18 months as the design has evolved 19 and the prospects and the reality of needing a demonstration i

20 project for CLEAR, GCRA with input and support from all the i

21 other program participants has been underway with a j 22 ' demonstration project initiative. We have spent probably l

23 collectively over a couple million dollars of private sector

) 24 funds for this effort and the effort for the demonstration 25 project definition study, and a resulting strategy plan.

I

65 i 1 The objectives of such a pr'oject are summarized 1-p

(,, 2 here. Fundamentally, for any nuclear demonstration there's 3 concern on performance, cost, schedule, being able to 4 establish a vendor and a utility constituent customer market.

! 5 But unique to the modular gas-cooled reactor is demonstration 6 associated with the licensing process, which we hold paramount 3

7 to this list of objectives that are summarized.there.

8 So there's a very fundamental process here through 9 this project to realize the advantages of the modular 10 gas-cooled reactor vis a vis the licensing process. And in 11 addition, to take advantage of this demonstration project and

(

12 its support for the certification process that we're i 13 ultimately targeting the design process to.

i 14 (Slide.]

i 15 The study that I referred to had the following 16 outline which I'll just touch on the key points. Another key 17 factor of interest on the modular concept is the ability to 18 build one module with the full-sized associated support 19 facilities for the nuclear island; that will then be the full 20 demonstration for a reference plant that would involve l

21 multiple modules. So a one-module plant allows a major i

22 reduction in the associated demonstration cost that would be p 23 needed for this technology.

l l 24 We have studied various licensing approaches, all 25 the way from an unlicarsed facility to a fully licensed l

T 66 1 facility. We opted for the latter 1:n the sense of fulfilling

\ 2 that major demonstration objective we touched on; namely, of 3 demonstrating that you can efficiently license such a module.

4 We looked at several siting optio.;s. We've

5 established the INEL site as our reference with regard to the

! 6 flexibility that that site offers. We're doing the kind of 7 testing that a remote site allows for having a very supportive 8 community for a nuclear test project per se.

j 9 A two-year test program has been laid out as a i -

10 preliminary basis for the kind of test that would be done for 11 performance for the capability of doing special tests with 12 regard to investment protection; certification support type

13 tests that this facility would allow and support the 14 regulatory process per se.

15 of course, the technology requirements as part of i 16 the cost; we need to qualify the fuel, finish the materials 17 qualification, et cetera. We've addressed the cost. We think 18 this total effort, total including the requirements for the i

19 technology, two years of demonstration operation, the fuel 20 supply, et al would be on the order of $800 million. There's 21 a chart in your handout that gives some further breakdown of i 22 these costs. of course, they are at a stage of conceptual 23 design and it's a planning ballpark number to be viewed at

• 24 this point. And a pregram schedule which of course is driven

25 by the resources that are available here on the front end, 4

67 1 which I'll come back to.

O2 .

COMMISSIONER ASSELSTINE: The demo plant would be 3 one module?

4 MR. MEARS: Yes.

5 [ Slide.]

6 Pertinent to the schedule, the strategy that has 7 evolved in parallel with the definition study envisions 8 basically a four-phased program whereby we can collectively --

9 we in the national sense -- address the key milestones 10 pertinent to each one of these phases as a basis for going on 11 to the next. We are into this project definition phase and 12 feel we have been for the last 18 months. We've projected 13 through the end of fiscal 1988. Key to the process, as you'll 14 see on the chart in more detail, is the NRC's license ability 15 statement to bona fide the ability to accrue the savings and 16 the efficiencies that we believe this unique concept affords.

17 Following that, we'd be into the preliminary design 18 and licensing, looking toward the submittal of a PSSAR for the 19 reference plant, and h PSAR specifically for the demonstration 20 project on the specific site. That will lead, of course, down 21 to the detailed design and licensing, we'd be into long lead 22 material ordering, as that precedes the construction start, 23 and after that is the final commitment for the major 24 manufacturing, construction, testing which would then be the 25 basis for commercialization.

3 68

1 [ Slide.]

2 That's a bit busy for your view here but you have it 3 J.n your handout. It's a chart that summarizes the key steps l 4 of the overall program. The first line is the ongoing 5 reference plant development activity and its pre-application 6 process which we are into today with the NRC Staff. And you 7 will note the PSID has been submitted. We are looking toward i

8 this NRC licenseability statement around the first quarter of 9 fiscal 1988.

4 10 The technology program of course is driven by that 11 design so that the design sets the needs and the requirements

12 for the technology program.

13 The fourth bar addresses the effort that I had-l 14 mentioned with regard to the definition and the execution of a 15 demonstration project, leading toward the PSAR by the and of 16 1990 and a construction start we hope by 1992. Startup in the 17 timeframe of 1995, a two-year period for demonstration tests L

i 1

18 which will then be the input to the final SSAR. We hope the 19 FDA and the certification for commercial plants there to

20 follow.

21 If successful, the chart here would show that the l

i j 22 first commercial plant then would proceed with a one-stop j 23 construction permit and operating license, and we could see 1

l s

24 the reality of this effort by the turn of the century.

25 In closing, let me say that in paral'.el with this i

i

. _ _ _ _ . _ _ _ _ . . _ . _ . _ _ _ . - ~ _ . _ . _ . . . . . . _ _ . _ _ . _ . . _ _ _ _ . _ . _ . , _ _ _ . _

.. . . . . _ - - . -_ .. .- - . . ~. . .. . - - - _

T 69 1 effort, that being the project defin'ition study within GCRA 1

i s,

s 2 and the other program participants, we have had underway at 3 GCRA an effort to produce a project strategy plan. This 4 effort addresses the organizational development, the l

I 5 management arrangements, the cost-risk sharing arrangements if f

6 you will for the' execution of such project.

i 7 Raising the utility support required for that effort

! 8 will be a formidable challenge. The most critical near term 9 impact that will address the feasibility for such an effort 10 being worth of utility support will be the process with and 11 response from the NRC on this near term activity of a PSID 12 review, an SER, and a licensability statement.

13 We strongly believe that the modular gas cooled

14 reactor and its unique fuel particle concept of containment i 15 offers unique advantages for the nation as a superior second 16 generation of nuclear power. At the same time, it requires a e

17 major challenge of the resources of the nation and the parties 3

18 involved to make sure we accommor 'a the unique advantages l 19 within the licensing process.

20 It is very clear that the feasibility of achieving 21 the commercial reality of the concept will, in large part, 22 depend on being able to accrue those unique advantages within 23 this licensing process and we look forward to the continued I

24 time and attention of the Commission to that end.

I 25 Thank you.

i

- . . . _ . - - . _ _ _ , . . _ - . . . _ _ _ . _ _ _ _ . . - - . . . , , _ . ~ . _ _ _ _ _ _ . , _ . _ , _ _ _ _ _ _ . _ - - , _ _ , , _ _ _ _ _ - . , _ _ _ _ _ .

1 70 1 CHAIRMAN ZECH: Thank you v'ery much.

2 Dr. Rossin, do you want to come back and join us and 3 perhaps I will ask my fellow commissioners for their 4 questions. And I will ask, in a few moments, for Commissioner 5 Roberts -- if we go on -- to take over the chair from me.

l 6 But first of all, let me ask my fellow commissioners 7 for their questions.

8 COMMISSIONER ASSELSTINE: Lando, I had a few, but

, 9 rather than rush through them at this point, maybe what we l

i 10 might coesider is getting the staff to sort of outline, in a i 11 brief paper for us, what their process is for looking at all 4

i 12 of these things, what some of the more significant policy 13 issues are that they see and then maybe we could get together 14 again, at some point, and discuss with them and perhaps 15 Mr. Rossin and some of the other participants in more detail, 16 given sort of the lateness of the hour now.

17 I sort of suggest that we think about doing 18 comething like that.

19 CHAIRMAN ZECH: I think that's a good idea.

20 Dr. Rossin, does that give you a problem at all?

21 MR. ROSSIN: Not at all. I think the presenters 22 that you heard this morning have indicated an enthusiask and a 23 willingness to go to great lengths to get their points across 24 to you. And I think they'd appreciate the opportunity to 25 explore this in more detail. You've seen presentations 1

71 1 summarizing the design and the approach this morning. And I

/ 2 think you will have substantive questions and I think they'd 3 like to answer in substantive fashion.

4 So if we could schedule a session and include staff 5 input to your questions, I think that would be fine. Perhaps 6 just a closing remark.

7 I think the presentations indicate a response to a 8 challenge and also a response to your advanced reactor policy 9 statement. Within these three different concept 10 presentations, I think you saw some similar thinking, 11 addressing key points. They're of extreme importance if we're 12 going to have an alternative, if we're going to have more than 13 one approach to solving our energy problems in the nuclear 14 area.

15 I think, just as an indication, you saw the 16 importance of licensability and of the approach to licensing 17 in all of these presentations. And what that tells you is 18 that if we're going to succeed, that licensability is a key 19 part of this. We're going to have to work together. We're 20 going to have to depend heavily on your commitment, as well as 21 our own, to get from here to there.

22 CHAIRMAN ZECH: Thank you very much. And I think we 23 should have another meeting. And I would suggest that the 24 meeting that has been proposed would be with the staff 25 presenting to us their comments and any recommendations or

72

, 1 proposals they may have on these matters. And then I would 2 ask, Dr. Rossin, if the presenters here could be at that 3 session to answer any questions the commissioners may have.

4 We'd be happy to have you come back, too, if it's 5 convenient for you, at that time. And we've been able to 6 pursue.some of these things in some depth. I agree that all 7 the presentations presented certainly have been a responsible 8 and forward-looking approach to many of the issues that we-9 discussed and the advanced reactor policy and the severe 10 accident policy and other important programs that we've 11 addressed here in the past year.

12 So I appreciate very much the thinking and the i

, 13 responsible way that these organizations have attempted to 2

\

14 look to the future.

15 And cartainly we recognize that at the Commission 16 our responsibility is to work closely on behalf of the 17 American people for the insured safety of reactor operation in l 18 our country. And we look forward to working with the 19 Department of Energy and with the other organizations that 20 presented here today.

I 21 I think a future meeting would be most appropriate 22 and we will certainly schedule that at a time that's i

23 convenient to you and to the other organizations at present.

24 So with that let me -- unless my fellow 25 commissioners have any final remarks to make, let me thank all i

73 1 of you for your very thoughtful and' responsible presentations 2 this morning. I think that they have been very valuable. We 3 will look forward to the next session when we can perhaps 4 explore in more depth with our staff, and with you and others, 5 the issues that we have discussed this morning.

6 Thank you very much and this meeting is adjourned.

7 [Whereupon, at 11:49 a.m.,.the meeting was 8 adjourned.]

9 10 11 12 13 l 14 15 16 1

17 18 19 20 21 22 23 24 25 1

l 1

% 2 REPORTER'S CERTIFICATE 3

4 This is to certify that the attached events of a 5 meeting of the U.S. Nuclear Regulatory Commission entitled:

6 7 TITLE OF MEETING: Briefing on Advanced Reactor Designs (Public Meeting).

8 PLACE OF MEETING: Washington, D.C.

9 DATE OF MEETING: Thursday, October 9, 1986 10 11 were held as herein appears, and that this is the original 12 transcript thereof for the file of the Commission taken x 13 stenographically by me, thereafter reduced to typewriting by

\

14 me or under the direction of the court reporting company, and 15 that the transcript is a true and accurate record of the 16 foregoing events.

17 18 --------------------------------

Pamel, Briggle 19 20 21 22 Ann Riley & Associates, Ltd.

23 24

10/8/86 SCHEDULING NOTES TITLE: BRIEFING ON ADVANCED REACTOR DESIGNS

/'~' ,

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SCHEDULED: 10:00 A.M., THURSDAY, CCTOBER 9, 1986 (OPEN)

DUPATION

APPROX 2 HRS 10 MINS PARTICIPANTS: DEPARTMENT OF ENERGY

- DR. A. DAVID ROSSIN ASST. SECRETARY FOR NUCLEAR ENERGY ROCKWELL INTERNATIONAL (LIOUID METAL DESIGN) 20 MINS

- JOHN S. MCDONALD DIVIS!ON DIRECTOR, Al

- ROBERT T. LANCET, PROGRAM MANAGER i

REACTOR PLANT SAFETY & RELIABILITY, AI

- CHARLES L. STORRS, COMBUSTION ENGINEERING i DIRECTOR OF HTGR/LMR DEVELOPMENT GENERAL ELECTRIC (l!OUID METAL DESIGM) 20 Mths l

, - J.S. ARMlJO MANAGER OF NUCLEAR SYSTEMS AND TECHNOLOGY OPERATICN

- GLENN SHERWOOD MANAGER OF LICENSING GA TECHNOLOGIES (HIGH TEMPERATURE GAS DESIGN) 20 MINS

- RICHARD DEAN I VICE PRESIDENT, GA TECHNCLOGIES

- T.E. NORTHRUP, GENERAL MANAGER POWER REACTOR PRCGRAMS, GA TECHMCLOGIES I

- WARREN CHERNCCK VICE PPESIDENT, CCMBUSTION EhGINEERING

- DANIEL MEARS, GAS COOLED REACTCR ASSOCIATES l

Hl/IID-BD 64 i

PRELIMINARY DRAFT SAFR SODIUM ADVANCED FAST REACTOR PRESENTATION TO NUCLEAR REGULATORY COMMISSIONERS i WASHINGTON, D.C.

OCTOBER 1986 l

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O O O I

i AGENDA e INTRODUCTION ,

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• DESIGN OVERVIEW I

e SAFETY GOALS / FEATURES l

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• ECONOMICS I

e FIRST PLANT PLAN -

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l e ASSESS ELECTRIC INDUSTRY NEEDS e DEVELOP SAFR TO MEET THESE NEEDS e ESTABLISH BASIS / APPROACH FOR DESIGN l CERTIFICATION LEADING TOWARDS COM-

! MERCIAL INTRODUCTION i

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O O O 1

i SAFR PROGRAM TEAM

) THE SODIUM ADVANCED FAST REACTOR (SAFR) TEAM MEMBERS AND THEIR RESPONSIBILITIES ARE SHOWN IN THE UPPER PORTION OF THIS CHART. THIS

OUTSTANDING GROUP HAS COLLABORATED FOR MANY YEARS ON LMR WORK,

! WITH EXCELLENT RESULTS. DOE HAS ORIENTED THE LMR BASE PROGRAM TO i PROVIDE STRONG SUPPORT TO THE INNOVATIVE DESIGN EFFORTS. WE HAVE DEVELOPED A DETAILED, COORDINATED SET OF WORK PLANS WITH OTHER BASE PROGRAM ORGANIZATIONS, AND THE RESULTS ARE SHOWN IN THE BOT-TOM PORTION OF THIS CHART. THIS- SECTION SHOWS THE NUMEROUS TASK AREAS IN WHICH BASE PROGRAM WORK IS BEING DONE BY LABORATORIES AND OTHER CONTRACTORS FOR SPECIFIC INCORPORATION IN THE SAFR DESIGN PACKAGE.

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THE SUPERIOR FUNDAMENTAL PROPERTIES OF LIQUID METAL REACTORS TRANSLATE INTO VERY FAVORABLE LMR PLANT CHARACTERISTICS. BY CAPI-TAllZING ON THESE PHYSICAL PROPERTIES, LMRs CAN BE DESIGNED SUCH j THAT NATURAL FORCES (E.G., GRAVITY, THERMAL EXPANSION) ESTABLISH SAFE AND STABLE CONDITIONS FOLLOWING ANY CREDIBLE REACTOR UPSET OR ACCIDENT. ADDITIONALLY, THE PROPERTIES OF SODIUM ALLOW PLANT EFFl-CIENCIES TO APPRAOCH THOSE OF FOSSIL PLANTS AND PERMIT USE OF CON-VENTIONAL HIGH SPEED STEAM TURBINES.

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! FUNDAMENTAL PROPERTIES OF LIQUID METAL REACTORS TRANSLATE INTO DESIGN ADVANTAGES b ~

PLANT CHARACTERISTICS COOLANT PROPERTIES EFFICIENT, LOW PRESSURE

HIGH THERMAL . COOLANT SYSTEM i CONDUCTIVITY j HIGH PLANT EFFICIENCY ,

1 LOW VAPOR PRESSURE CONVENTIONAL FOSSIL NONCORROSIVE OF STEAM TURBINES STRUCTURAL MATERIALS N PASSIVE DECAY HEAT

! HIGH BOILING POINT REMOVAL STRONG NATURAL LONG " GRACE PERIODS" CONVECTION

! O LOW RADIATION EXPOSURES

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LIQUID METAL REACTORS AROUND THE WORLD THIS MONTAGE PICTURES SEVERAL OF THE WORLD'S SODIUM-COOLED REAC-

) TORS. THE SODIUM REACTOR EXPERIMENT (SRE) (DESIGNED / CONSTRUCTED /

! OPERATED BY ROCKWELL INTERNATIONAL) WAS THE FIRST REACTOR IN THE WORLD TO SUPPLY ELECTRICITY TO A UTILITY GRID. THE 20-MWe EXPERI-MENTAL BREEDER REACTOR-il HAS OPERATED AT HIGH CAPACITY FACTORS FOR OVER 20 YEARS, AND THE FAST FLUX TEST FACILITY HAS OPERATED WELL

SINCE ITS STARTUP IN 1982. THE 250-MWe PHENIX, THE 1200-MWe SUPERPHENIX,

! AND THE 250-MWe PROTOTYPE FAST REACTOR (PFR) ARE OPERATING SUC-I CESSFULLY ON FRENCH AND UNITED KINGDOM UTILITY SYSTEMS.THE RUSSIAN l BN 350 HAS BEEN GENERATING ELECTRICITY AND PRODUCING DESALINIZED WATER FOR OVER A DECADE. ANOTHER RUSSIAN PLANT, THE 600-MWe BN 600, WENT CRITICAL IN JANUARY 1982 AND WAS TIED INTO THE GRID TWO MONTHS LATER.

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YEARS OF TEMPERATURE POWER

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BOR-60 U.S.S.R. 17 1110 OXIDE 60

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j BN-600 U.S.S.R. ,7 ~990 OXIDE 1470 PHENIX FRANCE 13 1040 OXIDE 567 l

SUPER PHENIX FRANCE 1 1013 OXIDE 2000 JOYO JAPAN 9 932 OXIDE 100

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EBR-il 23 882 METAL 62.5 l

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e NUMBER OPERATED (SOON 24 MORE) 61 j

• LARGEST UNIT 375 MW (SUPER PHENIX)

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• NUMBER.lHX OPERATING YEARS 400

• IHX AVAILABILITY >95%

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• LONGEST SINGLE IHX SERVICE TO DATE 20 YEARS (EBR-II) e SODIUM-HEATED STEAM GENERATORS

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• NUMBER UNITS OPERATED 750 j

• LARGEST UNIT OPERATED 750 MWt (SUPER PHENIX) i '

e STEAM GENERATOR PERFORMANCE 30-YEAR EXPERIENCE SHOWS AREAS REQUIR-ING SPECIAL ATTENTION

• LONGEST SINGLE STEAM GENERATOR l y SERVICE TO DATE -

20 YEARS (EBR-ll)

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O O O LMR PUMP EXPERIENCE

• l4 e MECHANICAL PUMPS i

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• NUMBER OPERATED 112 i e LARGEST UNIT 53,000 'gpm (HENGELO) j'

• NUMBER PUMP OPERATING YEARS 1,040

• PUMP AVAILABILITY EXCELLENT"

• LONGEST SINGLE PUMP SERVICE TO DATE 20 YEARS (EBR-II)

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• EM PUMPS

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• NUMBER OPERATED 70 (NOT INCLUDING AUXILIARY AND TEST PUMPS)

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• NUMBER PUMP OPERATING YEARS 300

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• PUMP AVAILABILITY EXCELLENT * -

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• LONGEST SINGLE PUMP SERVICE TO DATE 20 YEARS (EBR-il)

! *AS OF 1984 "ALMOST ALL PROBLEMS OCCURRED DURING TESTS AND COMMISSION-g ING PERIODS

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j e NUMBER OXIDE PINS IRRADIATED TO DATE 500,000

e OVERALL FAILURE RATE (WORLDWIDE) 0.01 %

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• PEAK BURNUP ACHIEVED IN PLANTS 100,000 mwd /T j e PEAK BURNUP ACHIEVED IN TEST PINS -

200,000 mwd /T e PEAK (NORMAL) MIDWAll CLAD TEMPERATURE 1290 F j e METAL FUEL PIN EXPERIENCE e NUMBER FUEL PINS IRRADIATED TO DATE 155,000 j e PEAK BURNUP ROUTINELY ACHIEVED IN PLANT 85,000 mwd j e PEAK BURNUP ACHIEVED IN TEST PINS 185,000 mwd l e PEAK (NORMAL) MIDWALL CLAD TEMPERATURE 1090 F j *AS OF 1984 1 n 4 to i

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O O O SAFR PROGRAM LOGIC j THEINITIAL SAFR DESIGN CONCEPT IS BEING DEVELOPED AND IMPROVED OVER i THE COURSE OF THE 37-MONTH PROGRAM, BASED ON TEST AND ANALYTICAL

! RESULTS OBTAINED FROM THE BASE LMR TECHNOLOGY PROGRAM, USER l NEEDS AS INDICATED BY UTILITY INTERACTIONS, FEEDBACK FROM NRC DIS-

. CUSSIONS, AND GUIDANCE RECEIVED FROM A" BLUE RIBBON" ADVISORY COM-l MITTEE.THE LATTER COMMITTEE WAS FORMED TO AID THE TEAM IN ADDRESS-ING THE VARIOUS INSTITUTIONAL ISSUES CONFRONTING THE PROGRAM. THIS

PROGRAM WILL RESULT IN A STANDARD DESIGN TAILORED TO MEET USER'S

! NEEDS AND IN PLANS AND COST / SCHEDULE ESTIMATES WHICH WILL PROVIDE A BASIS FOR PROCEEDING WITH THE EVENTUAL COMMERCIAL INTRODUCTION OF

) THE CONCEPT.

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O O O RETURN TO THE NUCLEAR OPTION It is just a matter of time before the United States must begin again to build new nuclear power i plants. With the demand for electrical power increasing at a rate of 2.2% per year, it is inevitable that the need for additional generating capacity, along with the cost savings provided by nuclear power, will bring the U.S. electrical utilities back to the nuclear option. Nuclear power still represents our best hope for providing electrical power through the next century.

However, before the next generation of nuclear power plants are built, the problems which plagued the first generation must be solved. These problems, which have contributed to the hiatus on orders for nuclear power plants, involve the three areas of: safety, operations, and economics.

Although today's light water cooled reactor (LWR) power plants do not present a real hazard to the general public, the next generation of plant must be farsafer. The Three Mile Island accident and now the Chernobyl disaster have created a public perception that all nuclear power plants

~

constitute a grave safety hazard.The advent of ariinherently safe reactor concept will do much to regain the public confidence and will also protect the utility's capital investment, facilitate licensing, and reduce the cost of operation. ,

Tomorrow's reactor power plants must be less complicated and more " operator friendly." The complexity of today's plants have led to lower than desirable plant availability performance and to operator mistakes, some with very expensive consequences. Also, the radiation exposure of operating personnel must be reduced in order to meet continually tightening standards and l reduce the cosi of replacing personnel who have been " burned out."

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g S Most important, tomorrow's reactor plants are going to have to be less expensive, both in ter ms of 8 initialinvestment as well as life cycle costs,if they are to be a viable alternative to coal-fired power

@ plants.

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84 85 86 87 88 89 90 91 92 93 94 YEAR I 4 DATA FROM

1985 ELECTRIC POWER SUPPLY AND DEMAND E NORTH AMERICAN ELECTRIC RELIABILITY COUNCIL 5 . .

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O O O l

TRANSLATING MARKET NEEDS INTO SAFR GOALS

! THE GOALS SUMMARIZED ON THE NEXT TWO PAGES WERE ESTABLISHED FOR j SAFR AFTER ANALYZING CAPACITY PROJECTIONS AND THE COMPLEX OF CON-

, CERNS WHICH TODAY CONFRONT THE UTILITY INDUSTRY. ACHIEVEMENT OF 1 l THESE GOALS IN A PLANT DESIGN SHOULD GO FAR TOWARDS ALLEVIATING UTILITY CONCERNS AND PROVIDING A VALUABLE ELECTRICAL POWER OPTION. ,

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SAFR GOALS l

• ASSURED PUBLIC PROTECTION

• ESTABLISH SAFE CONDITIONS WITH l NATURAL FORCES

• AVOID NEED FOR FAST OPERATOR

• ACTION

• OBVIATE NEED FOR EVACUATION l e ASSURED INVESTMENT PROTECTION

• LOW PLANT DAMAGE PROBABILITY e EXCEED ASME-SC-C <10~*/ YEAR

• LOW ACCIDENT-CAUSED e EXCEED'ASME-SC-B <10 / YEAR

, DOWNTIME e MINIMUM DEVELOPMENT RISK AND e USE EXISTING TECHNOLOGY; DESIGN

COST COST-EFFECTIVE TEST PROGRAM i

e HIGH CAPACITY FACTOR e >80%

• LONG PLANT LIFE e >60 YEARS

• LOW PERSONNEL RADIATION e <25 MAN-R/ YEAR Q EXPOSURE

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O O O i

i SAFR GOALS -

l l e COMPETITIVE ENERGY COSTS * <50 mills /kWh

• SHORT ONSITE CONSTRUCTION e <4 YEARS l SCHEDULE i

l 8 WIDELY USABLE STANDARD PLANT - e >75% OF U.S. SITES i

e LIMITED FINANCIAL RISK e MINIMlZE TIME TO INCLUSION OF INVESTMENT IN RATE BASE j e MINIMlZE UNCERTAINTY IN PRE-DICTION OF CAPACITY REQUIRE-l MENTS/ PLANT COST, SCHEDULE i

j e CLOSELY MATCH CAPACITY ADDI-TlONS TO GROWTH DEMANDS e MINIMlZE COMPLEXITY OF MANAG-

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O . O O -

WITH SAFR, SAFETY HAS NUMBER ONE PRIORITY SAFR is inherently safe, needing no protection systems or operator actions to protect the public and the utility's investment from the elfects of design basis accidents. This is the ultimate safety feature of SAFR, though protection system provide reliable reactor protection under normal circumstances.

Redundant and independent shutdown methods assure reactor transients are terminated and core cooling af ter shutdown is provided by passive systems which do not depend upon 'he action of any valves, switches, or actuators.

Much of SAFR's inherent safety is provided by the liquid sodium coolant, whose high saturation temperature (1630*F) and heat capacity allow use of natural convection cooling and provide opera-tors with long " grace periods" for reacting to accident conditions.

Utilizing such a safe reactor design means the expensive and complicated emergency injection and

cooling systems required for LWRs can be done away with, saving millions of dollars in plant cost and significantly reducing plant complexity. Eliminating these systems also increases plant availability by eliminating numerous sources of component failures which could cause plant shutdown.

An inherently safe reactor means easier licensing and increased public acceptance. Since SAFR will be a standardized plant design, able to be sited throughout the United States, obtaining construction and operating licenses will be faster and easier than has been the case in the past.

Because SAFR can withstand the worst conceivable accident and not sustain damage to the core or g the rest of the reactor, the possibility of losing the entire plant investment in a "Three Mile Island" S type event is eliminated. Also eliminated is the risk of incurring financialliability for the health and 4 safety of nearby residents.

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j e INHERENT RESPONSE TO ALL CREDIBLE EVENTS e SAFE CORE RESPONSE WITHOUT SCRAM

• PASSIVE DECAY HEAT REMOVAL e PASSIVE RESPONSE TO STRUCTURAL FAILURES *

] e MINIMlZE POTENTIAL FOR SEVERE ACCIDENTS l

j e RELIABLE DIVERSE SHUTDOWN SYSTEMS j e FEW CHALLENGES TO PLANT TRIP SYSTEM j e MINIMUM DEPENDENCY ON SUPPORTING SAFETY EQUIPMENT i

e LONG TIME PERIODS FOR CORRECTIVE ACTION e ELIMINATE NEED FOR EVACUATION i e DEFENSE IN-DEPTH WITH MULTIPLE BARRIERS l

e LOW LEAKAGE CONTAINMENT

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e LOW CORE MELT PROBABILITY j e LOW PLANT RISK i E b

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O O O

. MAJOR FEATURE SELECTIONS l THE MAJOR FEATURES OF SAFR WERE CHOSEN AFTER THE TEAM, IN CLOSE

! COOPERATION WITH ARGONNE NATIONAL LABORATORY AND CHICAGO BRIDGE AND IRON, CONDUCTED A COMPREHENSIVE, DETAILED REEVALUATION OF THE l

OPTIONS AVAILABLE.THIS CHART INDICATES VARIOUS OPTIONS CONSIDERED FOR SOME OF THE MAJOR FEATURES. IN PARTICULAR, THE 350-MWe MODULE i SIZE, CHOSEN AFTER A LENGTHY AND DETAILED STUDY, INDICATED THAT THIS i

SIZE OFFERS SIGNIFICANT COST ADVANTAGES OVER SHIPPABLE SMALLER

UNITS, AS WELL AS AN OPTIMUM BALANCE OF FACTORS, PROVID!NG i

j e COST, SCHEDULE, QUALITY ADVANTAGES OF SHOP FABRICATION

• ECONOMY OF SCALE e MINIMUM FIELD CONSTRUCTION AND TIME TO STARTUP

• DIVERSE, PASSIVE SHUTDOWN HEAT REMOVAL SYSTEMS l

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O O O MAJOR FEATURE SELECTIONS SAFR CHARACTERISTICS 9

FEATURE OPTIONS EVALUATED BEST CHOICE 9 PLANT SIZE / 9 1000-1350 MWe MONOLITHIC (FIELD 9 350 MWe BARGE SHIPPABLE ARRANGEMENT FABRICATION) MODULAR PLANT -Si(ARED G RAll AND BARGE SHIPPABLE FACILITIES 9 110-350 MWe MODULES 9 NSSS '

O POOL, LOOP G POOL CONFIGURATION O FUEL TYPE 9 OX1DE, METAL 9 INHERENTLY SAFE METAL CORE G STEAM CYCLE O SATURATED STEAM -

G 950"F REACTOR OUTLET, 9 SUPERHEATED STEAM - SULZER CYCLE SUPERHEATED BENSON 9 SUPERHEATED STEAM - BENSON CYCLE STEAM CYCLE 9 SAFETY SYSTEMS 9 REACTOR SHUTDOWN - DIVERSE, REDUNDANT, .

O PASSIVE, LOCAllZED SYS-SELF-ACTUATING TEMS THROUGHOUT 4 SHUTDOWN COOLING - FORCED AND NATURAL CONVECTION, VARIOUS LOCATIONS IN HEAT TRANSPORT PATH 5

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THIS SHOWS THE FLOW DI AGRAM FOR THE 350-MWe SAFR STANDARD UNIT. THIS

DESIGN IS BEING DEVELOPED WITH DESIGN EMPHASIS ON (1) SIMPLICITY AND

LOW CONSTRUCTION QUANTITIES, (2) PASSIVE SAFETY FEATURES, (3) MINI-MlZING AND LOCALIZING THE SYSTEMS / COMPONENTS UPON WHICH NUCLEAR SAFETY DEPENDS, AND (4) " USER FRIENDLINESS." THE DESIGN IS BASED ENTIRELY ON EXISTING TECHNOLOGY. THE REACTOR ASSEMBLY IS SHOP-FABRICATED AND SHIPPED TO THE SITE. ESSENTI ALLY ALL RADIOACTIVE MATE-RIAL IN THE PLANT IS CONTAINED WITHIN THE 39-ft-DIAMETER POOL-TYPE REACTOR VESSEL. THE DIAGRAM ALSO SHOWS FEATURES INVOLVED IN THE DIVERSE, PASSIVE DECAY HEAT REMOVAL SYSTEMS, AS WELL AS THE CONVEN-TIONAL ONCE THROUGH SUPERHEATED STEAM CYCLE EQUIPMENT. UNITS SUCH AS THIS CAN BE USED IN MULTIPLES AS NEEDED TO PROVIDE THE REQUIRED GENERATING CAPACITY.

4 -

Rockwelllntemational seansesteen)susussamme Rocketdyne Division 15-35-0

O O O 350 MWe STANDARD SAFR UNIT FLOW DIAGRAM

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THIS ARTIST'S CONCEPT ILLUSTRATES THE USE OF MULTIPLE STANDARD 350-MWe UNITS - IN THIS CASE FOUR " MODULES" WHICH COMBINE Td PROVIDE 1400-MWe GENERATING CAPACITY. 350-MWe MODULE SIZE WAS FOUND TO BE THE OPTIMUM TRADEOFF OF SHOP-FABRICABILITY, LOW CAPITAL COST, PAS-SIVE DECAY HEAT REMOVAL SYSTEMS, SHORT CONSTRUCTION SCHEDULE, AND PRODUCT INVESTMENT RISK. THE TURBINE GENERATOR FACILITIES ARE SEPARATE FROM THE NUCLEAR ISLANDS AND CAN BE CONSTRUCTED USING NONNUCLEAR STANDARDS WITH CONCOMITANT COST SAVINGS.

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. SAFR EMPLOYS A POOL-TYPE REACTOR IN WHICH THE PRIMARY INDUCER

.. PUMPS AND INTERMEDIATE HEAT EXCHANGER ARE POSITIONED WITHIN THE REACTOR VESSEL AROUND THE CORE. ALL PRIMARY RADIOACTIVE SODIUM IS CONTAINED WITHIN THE SIMPLE CYLINDRICAL REACTOR VESSEL, WHICH HAS NO NOZZLES, ATTACHMENTS, OR PENETRATIONS. THE REACTOR INTERNALS AND DECK STRUCTURE ARE DESIGNED SUCH THATTHE COMPONENT PARTS OF THE ENTIRE REACTOR ASSEMBLY CAN BE INTEGRATED WITHOUT BIMETALLIC JOINTS. THE REACTOR VESSEL, WHICH IS FULLY INSPECTABLE, PROVIDES THE ONLY TENSION MEMBER IN THE CORE SUPPORT LOAD PATH. A SIMPLE SET OF CYLINDERS FORM THE ACTUAL CORE SUPPORT STRUCTURE. THE THERMAL CAPACITY PROVIDED BY THE MASS OF SODIUM IN THE POOL PROVIDES A LONG GRACE PERIOD BEFORE ANY OPERATOR ACTION MAY BE REQUIRED.

l 3

i Fk,ckwellintemational scenewstsess)susssssamasse 15-42-0 Rocketdyne Division

O O O

~

PASSIVE, DIVERSE NATURAL CONVECTION DECAY HEAT REMOVAL SYSTEMS THE THREE DIVERSE, INDEPENDENT METHODS OF DECAY HEAT REMOVAL -

NAMELY, THE NORMAL HEAT TRANSFER PATH,THE DIRECT REACTOR AUXILI ARY COOLING SYSTEM (DRACS), AND THE REACTOR AUXILIARY COOLING SYSTEM (RACS) - ASSURE THAT THE REACTOR WILL NOT EXPERIENCE EXCESSIVE TEMPERATURES.

! 5 FlockwellIntemational sceneesisese)ssesseessneess

, Rocketdyne Division 15-49-0 l

O O O SIMPLE POOL CONFIGURATION CONTAINS ALL PRIMARY COOLANT DECK

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DECAY HEAT ACCOMMODATED WITH MARGIN THIS PLOT OF HOT POOL TEMPERATURES FOLLOWING PLANT SHUTDOWN SHOWS THE ABILITY OF SAFR TO ASSURE ADEQUATE CORE COOLING, EVEN ,

WHEN THE LOSE OF NORMAL AND DRACS SYSTEMS IS POSTULATED. IN FACT, IF THE LOSS OF ALL COOLING IS POSTULATED, THE LARGE THERMAL INERTIA OF SAFR RESULTS IN A "LONG" GRACE PERIOD IN WHICH TO TAKE CORRECTIVE ACTION.

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LMRs CAN, BY CAPITAllZING ON SODIUM PROPERTIES, BE DESIGNED TO BE SELF-REGULATING / POWER LIMITING UNDER VERY SEVERE CONDITIONS. THIS

EFFECT IS ILLUSTRATED BY THE EXPERIMENTAL RESULTS SHOWN ON THE FOL-l LOWING THREE PAGES FROM THE EBR-ll AND FRENCH RAPSODIE REACTOR.

SAFR INCORPORATES THIS CAPABILITY.

3 i

I Rockwellintemattonal seasnesteese)essessemenssee Rocketdyne Divlelon 14-74-0 l

O O O -

l RAPSODIE REACTOR TRANSIENT - LOSS OF PUMPS WITHOUT SCRAM I i 25 SODIUM BOILING POIN'

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O O DIVERSE SELF-ACTUATED SHUTDOWN SYSTEM THE PRIMARY SHUTDOWN SYSTEM FOR SAFR IS THE WELL-TESTED CRBR SYS-TEM. THE DIVERSE, SECONDARY SYSTEM IS A SELF-ACTUATED CONCEPT WHICH HAS BEEN UNDER DEVELOPMENT FOR SOME TIME ON DOE's LMR BASE PRO-GRAM. IN THIS CONCEPT, ARTICULATED ABSORBER ELEMENTS ARE SUPPORTED ABOVE THE CORE BY A MAGNETIC LATCH.THE LATCH IS AN INTEGRATED CON-FIGURATION SUCH THAT INCREASES IN SODIUM EXIT TEMPERATURE FROM THE CORE ARE RAPIDLY COMMUNICATED TO THE LATCH. IF THE OVERTEMPERA-TURE IS SUFFICIENT, THE CURIE POINT OF THE LATCH MATERIAL IS REACHED AND THE ABSORBER IS RELEASED - DROPPING INTO THE CORE AND SHUTTING DOWN THE REACTOR. THE PLOT OF TEST DATA ON THE OPPOSITE CHART SHOWS THAT THIS DEVICE PROVIDES A GOOD MARGIN (EVEN UNDER '#ORST CASE TOLERANCE STACKUPS) FOR SAFE SHUTDOWN OF THE REACTOR.

h Flockwelllntemational sesenesteess)ssesessenesses Rocketdyne Division 15-44-0

O O O INDEPENDENT, DIVERSE, REDUNDANT SHUTDOWN SYSTEMS PRIMARY SECONDARY c,olt, wo NUMBER 9 3 f-ACTUATION PPS (2-OUT-OF-4) 1. MAGNETIC FLUX SWITCH / Na ._ MAGNETIC

?.-- tArcu OVERTEMPERATURE g 2.PPS RELEASE MECHANISM ROLLER NUT MAGNETIC LATCll I $sou INSERTION SPRING ASSISTED GRAVITY k GRAVITY CONTROL ASSEMBLY SINGLE RIGID ABSORBER ARTICULATED ABSORBER PERFORMANCE ACillEVES COLD SHUTDOWN ACHIEVES HOT SHUTDOWN DASitPOT WITH SINGLE FAILURE

• WITH SINGLE FAILURE * ' , /

s PROVEN DESIGN CRBRP, FFTF, LSPB LMFBR BASE PROGRAM - SASS INilERENCY FAIL-SAFE ON LOSS ACTUATES ON LOF, TOP Z.5_I {;

OF SERVICES gl gl gl u' 'u' @

q alNCLUDING STUCK ROD s;

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Rockwellinternational e g sonneusTson)susemesasus Hocheldyne Division g

O O O SAFR CONTAINMENT CONCEPT SECONDARY UPPER CONTAINMENT BOUNDARY g~

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fE W f: ::b: .oi 5." +;f I' % 79 flockwellInternational e

• g comenusnoNhENGINEERING b Rocketdyne Division g

l O O O PASSIVE SPENT FUEL COOLING tc s '.

POWER - KW INVESSEL STORAGE F/A B/A FUEL STORAGE MINIMUM 16.8 3.1 COOLING STACK MAXIMUM 25.4 12.4 /

TOTAL 551 150 m

/'

A-FRAME EXVESSEL STORAGE EXCHANGE MINIMUM 1.2 0.2 l 7

', MAXIMUM 1.8 0.9 - ..

,4. . .. [ _ _ _ _ _ _

TOTAL 40 11 ll - -

E.

/

',', / INVESSEL flap / g STORAGE ll

[f 4 l ..

s 81

/

• b  !

, l A-FRAME TO TDS - \ _t O

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EXCHANGES - 1 , ,j _

]

PEAK ASSEMBLY TEMPERATURES (*F) l F/A STUCK IN REACTOR PORT 463 F/A STUCK IN FTC F/A IN INTERIM STORAGE 319 /[e [ - ']

\

450 mI {- -- \

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)

i O

h FUEL STORAGE

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COOUNG AIR W IVS TO A-FRAME CORE TO IVS m EXCHANGES EXCHANGES

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l i

O- HOOF E XHAUST FANS FOR O

1 NORMAL VENTILATION.

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PANELS AS REQUIRED.

t ' IHTS PUMP 1 g , ,._

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l i- SWRPS

• lNSULATION l VAU LT

, l T No DRAINAGE TO SWRPS

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Q .VAU LT.

f Rockwellintemational  !

Rocheldyne Division seanowstsen)smeessemansa INSULATION

O O O

SUMMARY

OF SAFR INHERENT / DESIGN '

SAFETY FEATURES e INHERENT CORE " SHUTDOWN" FOR ALL CREDIBLE EVENTS EVEN IF TRIP SYSTEM DOESN'T FUNCTION

• PASSIVE DECAY HEAT REMOVAL SYSTEMS FUNdTION INDEFINITELY FOR i STATION BLACKOUT

• INHERENT SASS PROVIDES SHUTDOWN BACKUP AND SABOTAGE

] PROTECTION e CONTAINMENT VESSEL ASSURES CORE COOLING EVEN IF VESSEL LEAKS

• MULTIPLE PRIMARY PIPES PER PUMP ASSURES CORE COOLING EVEN FOR PIPE RUPTURE e IN-VESSEL FUEL STORAGE ASSURES PASSIVE FUEL CCOLING DURING ALL STORAGE AND ACCIDENT CONDITIONS e LARGE SODIUM INVENTORY PROVIDES INSENSITIVITY TO BOP TRANSIENTS AND LONG TIME FOR OPERATOR ACTION e FEW TRIP SIGNALS REDUCE CHALLENGES TO PLANT TRIP SYSTEM

• SECONDARY CONTAINMENT BACKUP OF PRIMARY CONTAINMENT REDUND-R ANT SEALS .

?

3

• PASSIVE SODIUM FIRE SUPPRESSION AND RACS INSENSITIVITY TO AERO-

$ SOLS ASSURES CONTINUED DECAY HEAT REMOVAL Rockwell kdemational flocketdyne Division 14-53-0

l O O O SAFR HAS LOW CORE MELT PROBABILITY l

FAILURE FREQUENCY (PER YEAR)

INTERNAL INCLUDING SHUTDOWN HEAT REMOVAL INITIATORS SEISMIC NHR ONLY 1.8 x 101 -

NHR AND DRACS 7.2 x 10-4 -

NHR AND RACS 1 x 10~8 - - -

NHR, DRACS AND RACS 3.6 x 10-s 1,1 x j o s*

REACTOR SHUTDOWN APTS ONLY 9.7 x 10-s APTS AND SASS 7.6 x 101 -

APTS, SASS AND CORE INHERENCY 7.6 x 10-" 5 x 10 a* * ,

• BASED ON GENERIC LWR FRAGILITY DATA m

• • BASED ON CONSERVATIVE ENGINEERING JUDGEMENT 8

3 5

8 m._ _.-

e RocketdyneOlvleton . .

l O O O l 10 LATENT RISK GOAL '

-(0.1% OF ALL CANCER FATALITIES)x PROMPT RISK GOAL -

O (0.1% OF ALL .

i 2 10 / / / l/////////

y-

/

{ DO O O

I _ O ,

b O W . %O I

$ 10 O l SAFETY o= WASu-i400 nESutTS i CA AI O j O= OTHER LWR HISK j uvMLO g - O ASSESSMENT RESULTS I

M TISIFi m gs " S " ^'"" 8 ^ '" " ' S

• ON 75TH PERCENTILE LWR a- "- "''""'^ "")

MTH$- @=1-UNIT SAFR PLANT I MARGIN g;35 -

2 . -

t b 10 @

oc SAFR lNTERNAL i

EVENTS AND SEISMIC 3

" 10-14 i i i i i . . >

a 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4

[ LATENT FATALITY RISK (PER YEAR, WITHIN 10 MILES)

, Rockwelllnternational seanmusnom)sucemusasus b Hocheldyne Division (g7

O O O SAFR STANDARD PLANT GOAL I

! e ACCOMMODATE 75% OF U.S. BARGEABLE SITES e SEISMIC LOADINGS e SOIL (FOUNDATION) CONDITIONS

! e METEOROLOGY i e HYDROLOGY

e TERRAIN j e DISTANCE TO POPULATION CENTERS e COOLING WATER 1

l e ACCOMMODATE UTILITY SYSTEM INTERFACE REQUIREMENTS e ACCOMMODATE ON-SITE OR OFF-SITE SPENT FUEL

REPROCESSING e PROVIDE COMPETITIVE NSSS AND BOP EQUIPMENT BIDS

, o PROVIDE COMPETITIVE CONSTRUCTION CONTRACT BIDS 0

em noch.wyn. ow.um

- g _ _ _ >.. _ ... _ 9-33-0

O O O TOP LEVEL SAFR STANDARDIZATION APPROACH I

e STANDARDIZE ALL SAFETY-RELATED SYSTEMS / COMPONENTS, E.G.,

i

• REACTOR ASSEMBLY

)

• CONTAINMENT BUILDING

! *RACS l

• PRIMARY PUMP

! e STANDARDIZE MAJOR PLANT INVESTMENT PROTECTION SYSTEMS /

) COMPONENTS, E.G.,

I *DRACS i

• SHUTDOWN SYSTEM i ePCS

!

• APTS .

e STANDARDIZE SODIUM SYSTEMS / COMPONENTS, E.G.,

I

• IHTS i e STEAM GENERATORS

• SECONDARY PUMPS ,

j e FUEL HANDLING SYSTEM e STANDARDIZE FOR PLANT ECONOMICS, E.G.,

• PLOT PLAN

• NI DETAILED ARRANGEMENT

@

• CONTROL BUILDING ARRANGEMENT 3

• NI MAINTENANCE BUILDING ARRANGEMENT 5

i E RockwellIntemational seassestson)asseessamesess Rochelde Deh 9-34-0

O O O SAFR PROMISES COMPETITIVE ECONOMICS THE PLOT SHOWS THE ENCOURAGING RESULTS OF THE DETAILED COST ESTl-MATES PREPARED BY THE SAFR TEAM IN ACCORDANCE WITH DOE GUIDELINES.

THESE RESULTS WERE CROSS CHECKED BY INDIVIDUAL TEAM MEMBERS BY AN INDEPENDENT GROUP, AND WERE RECONCILED WITH OTHER PLANT COSTS THROUGH DETAILED COMPARISONS OF FACTORS SUCH AS CONSTRUCTION QUANTITIES. OUR MAIN CONCLUSIONS ARE -

e THE FIRST SAFR PLANT MAY BE ENERGY COST-COMPETITIVE WITH COAL PLANTS e MATURE SAFR PLANTS CAN HAVE CAPITAL COSTS SIMILAR TO LWRs, AND ENERGY COSTS LOWER THAN LWRs a 1

? l 8 1 Rockwelllnternettonal 15-43-0 Flocketdyne Division

.I

O t O SAFR PROMISES COMPETITIVE ECONOMICS (1985 DOLLARS)

TO AWAY CYCLE '

C3 COLOCATED AND/OR '

CENTRAllZED REPROCESSING 70 -

.; $1100/kWe

$2  ;  !

$900/kWe

!Il!I i i e

60 uI p .. . miu ni;;mmummm d .

g l

FIRST PLANT j So -

COMMON FACILITIES s 350 MWe UNIT 40 - LWR $1300/kWe 1414/kWe E /

< 0 a

@ "O O $i248/kWe g 30 -

DOE GUIDELINES PROVIDED:

\0 ' *

] 20 -

O COAL AND LWR COSTS

@J G FINANCIAL FACTORS G CAPACITY FACTORS MATURE PLANTS

)

' 10 - a COAL AND LWR; 70%

s LMR; 80%

, m o I l l l 1995 2000 2005 2010

] ' OVERNIGHT COSTS E

Rockwellintemetkmal sensassison)sussasseuses i

Rocheldyne Divlelon

1 i

l -

SAFR FINANCIAL PLAN i

\

DETAILS TO BE PROVIDED LATER i

i

< {

(

i

, s 4

. T O

m Flockwellintemettonal Rocheldype Devleton seansestson)essessessesses I

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e . _ _ _ _______ _ __ - _ _ _ __ _ _ _ _ _ _ - _ _ _ _

! O O O 1

SAFR OFFERS HIGH POTENTIAL FOR i MEETING UTILITY NEEDS - l

't SAFETY GOALS MET WITH LARGE l MARGIN -

l

[

• SAFETY RESPONSE REQUIRED FOR PUBLIC

- INDEFINITE l SAFR GOALS l ,

• PRA INDICATES EVACUATION l PLANNING UNNEEDED 1

i e RESPONSE REQUIRED FOR j . .ssonco roooc raorection

. ggggs*lgoag,enous INVESTMENT PROTECTION - ~ 15 h

^ " '

! l 7.$$"[I.[o75^.".$o'"iU^ e e ASME-SC-C ~3 x 10~5/ YEAR DECAY

]

l

. ,ssonio.uvesimm eno .cm.

e LO.P ANT oAMAoE Pao 8A8ettiY . ENCEED AsME sC C<10 *WEAa

/ HEAT REMOVAL j . to..cc,oem causio oo.ui . . exce o s sc 8<io neaa

• ASME-SC-B <10 / YEAR DECAY HEAT i . -ooo o.mormm .s. '

REMOVAL

! . o.o..c.e.c.in cion ..>=

os exisnuo ric ot='\ ,

. tonoetam o,e . > n as e AVAILABLE ADVANCED TECHNOLOGY

. to. asonnu aio.^" "'"' $""'

• " "%""'^" N USED. FEATURE VERIFICATION TESTS j REQUIRED

! co .

e 85% ESTIMATED m

l

[ # ACHIEVABLE

) h

• 20 MAN-REM /YR ESTIMATED Rockwellintemational Rocketdyn. Division eennessison)esssensaasus

i j '

1 SAFR OFFERS HIGH POTENTIAL FOR i

MEETING UTILITY NEEDS - 11

• 39-48 mils /kWh (2nd UNIT) 29-38 mils /kWh (12th UNIT)

SM GORS 9 <3 YEARS ON SITE i

. ci. ...vt inino, cosis . < s. f* ACHIEVABLE - WSTERWAYS SERVE j ANTICIPATED MARKET

. sgigs.i cousinuci ou . .. vi.as

. ..oi t, us.m. si uo.no et. . . >rss o, o s siirs

1. <$600M MAXIMUM CASH OUTLAY FOR

. . in ein.ucio ais-

.g ,igiggoggigguo / MULTI-UNIT ORDER

• [$c'83ENh[a'"E'S -e STANDARD 350-MWe UNIT ACHIEVES j

• i=!','o"t=f,7E'!Ps . .- THESE

. g., g c.=u x , o, u. .o.uo . su-t., ocuo"

_ , , , , ,c,,,,, =* NUMBER OF SYSTEMS,1E POWER, "oa"

• ^" ,,, ,, , , \ RADWASTE DRAMATICALLY REDUCED -

s'ilh'u'nSsi!OEi s" FROM CRBRP/LWRs

• 8018'"'" ""'" "\

e 78% OF NRC-SCRUTINIZED WORK IN FIELD (LSPB) REDUCED TO 30%

9

• INCORPORATED T

, to h=

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g noca.w m

ya. oivi. ion

- g >._ -

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O O O i

1 KEY FEATURES REQUIRING DEMONSTRATION I

NON- LICENSABILITY I NUCLEAR TEST ISSUE TESTS TREAT EBR-Il FFTF (LT)

DRACS PERFORMANCE 1986 1985 COMPLETED 1997 RACS PERFORMANCE 1985/1987 1986 1997 SELF-ACTUATED SHUTDOWN 1976/1990 1987 1997 SYSTEM (SASS)

RBCB PERFORMANCE 1983/ 1997

1990 ANTICIPATED TRANSIENTS 1 WITHOUT SCRAM (ATWS)

EVENTS LOSS OF FLOW WITHOUT

! SCRAM 1989 1986 1986-1988 1997 LOSS OF HEAT SINK 1986 7 3

WITHOUT SCRAM 1997 -

y TRANSIENT OVERPOWER WITHOUT SCRAM 1989 1987 7 1997 l 8 l $

i s

! i l @ ==- @ -->--

% N ,

SAFR FORMAL NRC INTERACTIONS I.s l - l .,j l . l . I .,l ., . l . l .,l . l .,j . l l Il l. .

n " - - =.

Ed4AF T- - COL ES 1 NI

-I5IL CONSTRUCTIDH ej '

OPE IIDI SITE 'r A K) l ER ggggg3 l

I tlCENSADittIY  !

LETTER l SER  !

I COMMOra t,(W E.t l'I t at lite L 8. alt tAHW I4 5AR.

• l ACIMIIE S IIE 9GN I1Srila DEFK1N l tt PSE) SAffry l

TESI l Pt Ate g

l SAFR ACRS ST AnitfP IESinNG FOf t SID l ICATIOtl SER f, gra y s PL Atli FES

• FDA .g) nap r _

[S l

SI4ti)AnD ss 'r'r 'r 'r SI A3"I8 U E*E 'UI Pt Atat 88 l cot #SIHUCilurs /IESiwai(WtilAIK Wa

( ir <r j STD STD CONIRACT PtANI PI Atal IOlt N i SAll FSAll ll[ ACIOll

( ASSE MIN Y 7 Sale S 4 O Af s1 ER IlockwellInternational seassesnen)suseussesus flockoldyne Division g +

i

O O O l SAFR COMPLIES WITH ADVANCED REACTOR ACCIDENT POLICY STATEMENT PDLICY SAFR COMPLIANCE e COMPLIANCE WITH PROCEDURAL MEETS INTENT OF REGULATIONS REQUIREMENTS AND CRITERIA OF CURRENT COMMISSION REGULATIONS e DEMONSTRATION OF TECHNICAL COMPLIANCE INCLUDES ALL NUREG RESOLUTION OF ALL APPLICABLE 0933 ISSUES. DECAY HEAT REMOVAL UNRESOLVED SAFETY ISSUES AND SYSTEM RELIABILITY ASSURED AND MEDIUM-AND HIGH-PRIORITY ' INDEPENDENT OF AC AND DC POWER GENERIC SAFETY ISSUES e COMPLETE PRA AND CONSIDERATION PRA DRAFT SUBMITTED TO NRC OF SEVERE ACCIDENT OCTOBER 1985. NEXT UPDATE OCTOBER VULNERABILITIES 1986. REDUCTION OF SEVERE ACCIDENT PROBABILITY MAJOR SAFR GOAL'

• COMPLETE STAFF REVIEW WITH CON- PRELIMINARY STAFF REVIEW INITIATED

$ CLUSION OF SAFETY ACCEPTABILITY 1985 I

E Roc.kwellintemational sensmesteen)essousessnes nocheidyne osvisaan 14-47-1

O O O SAFR COMPLIES WITH ADVANCED REACTOR ACCIDENT POLICY STATEMENT I

i POLICY SAFR COMPLIANCE e COMPLIANCE WITH PROCEDURAL MEETS INTENT OF REGULATIONS REQUIREMENTS AND CRITERIA OF CURRENT COMMISSION REGULATIONS e DEMONSTRATION OFTECHNICAL COMPLIANCE INCLUDES ALL NUREG RESOLUTION OF ALL APPLICABLE 0933 ISSUES. DECAY HEAT REMOVAL UNRESOLVED SAFETY ISSUES AND SYSTEM RELIABILITY ASSURED AND MEDIUM-AND HIGH-PRIORITY INDEPENDENT OF AC AND DC POWER GENERIC SAFETY ISSUES e COMPLETE PRA AND CONSIDERATION PRA DRAFT SUBMITTED TO NRC OF SEVERE ACCIDENT OCTOBER 1985. NEXT UPDATE OCTOBER VULNERABILITIES 1986. REDUCTION OF SEVERE ACCIDENT PROBABILITY MAJOR SAFR GOAL i

e COMPLETE STAFF REVIEW WITH CON- PRELIMINARY STAFF REVIEW INITIATED y CLUSION OF SAFETY ACCEPTABILITY 1985 4

O

.idyn. o weson 14-47-1

! O o o

SUMMARY

OF SAFR DESIGN I

e INHERENT SAFETY CHARACTERISTICS ASSURE RISK AT LEAST AS LOW AS LWRs BUILT IN SAME TIME FRAME e ECONOMICALLY COMPETITIVE WITH COAL AND LWRs e DESIGN LENDS ITSELF TO PLANT-WIDE STANDARDIZATION e R&D BASE PLUS FIRST MODULE SAFETY TESTS SUFFICIENT FOR DESIGN CERTIFICATION i

e DEVELOPMENT SCHEDULE MEETS MARKET NEEDS i

e VIABLE FINANCIAL PLAN s

4 3

! b i s

~ ~

II.wr = b 14-48-1

O O O

) SAFR COMPLIES WITH ADVANCED REACTOR POLICY STATMENT

• AT LEAST SAME DEGREE OF PROTECTION OF PUBUC

• SAFR GOAL IS TO PROVIDE SAME DEGREE OF AND ENVIRONMENT AS CURRENT LWRs . PROTECTION AS CONTEMPORARY LWRs

• ENHANCED MARGINS OF SAFETY AND/OR SIMPUFIED e ENHANCED MARGINS OF SAFETY AND INHERENT

] INHERENT, PASSIVE OR OTHER INOVATIVE MEANS TO REACTOR SHUTDOWN AND DECAY HEAT REMOVAL ACCOMPUSH SAFETY

• COMPLY WiIH SAFETY GOAL POUCY

• SAFR EXCEEDS SAFETY GOALS IN NRC POLICY l

• HIGHLY RELIABLE LESS COMPLEX PREFERABLY

• INHERENT / PASSIVE SHUTDOWN AND DECAY HEAT INHERENT OR PASSIVE SHUTDOWN AND DECAY HEAT REMOVAL SYSTEMS

, REMOVAL SYSTEMS I

• LONGER TIME CONSTRAINTS AND APPROPRIATE

• INDEFINITE STATION BLACKOUT SHUTDOWN AND DECAY l lNSTRUMENTATION FOR OPERATOR ACTION HEAT REMOVAL.10s OF HRS EVENT WITHOUT SCRAM OR

DECAY HEAT REMOVAL 1

• SIMPLIFIED SAFETY SYSTEMS, REDUCED REQUIRED

• FEW/ SIMPLE SAFETY SYSTEMS. NO OPERATOR ACTION i

OPERATOR ACTIONS REQUIRED. 4 POWER PACK 1-E POWER ONLY 80 kW

!

• MINIMlZE POTENTIAL FOR SEVERE ACCIDENTS THROUGH

• INHERENT AND DIVERSE CORE SHUTDOWN AND DECAY

! INHERENCY, RELIABluTY, REDUNDANCY, DIVERSITY, HEAT REMOVAL FOR ALL CREDIBLE INITIATING EVENTS l lNDEPENDENCE

• REUABLE BOP

• RELIABLE BOP FOR PLANT INVESTMENT, REACTOR INSENSITIVE TO BOP '

• EASILY MAINTAINED e SUFFICIENT ACCESS PROVIDED e REDUCED RADIATION EXPOSURE * <25 MAN-R/ YEAR

• DEFENSE IN-DEPTH

• MULTIPLE RADIATION BARRIERS PROVIDED

• EXISTING TECHNOLOGY OR R&D COMMITMENT

• TECHNOLOGY DEMONSTRATED OR IN PLACE R&D

, PROGRAM 8

te l 4 .

7 O

Rockwellintemational seassesteen%essnessemos Rocketdyne Divleton

i l O O ""O-i i

I l

' SAFR I SODIUM ADVANCED FAST REACTOR I

PRESENTATION TO l NUCLEAR REGULATORY COMMISSIONERS WASHINGTON, D.C. .

g OCTOBER 1986 l

4 I

EOWBIBUSTIGOs ENGINEERIIGG RockwellInternational fo,"2.'=nOfi,co,"oeoes ";3 "

ee c'a: a l'.T.".

an a Pas Cahlosnia 91303 Ei?, Tea 7."st,"eti San Francisco,Cahtornia 94105 l

i i

l SAFR PROGRAM TEAM Rockwell

, 5 #jg International n.cs.w, ce e annesmes). sos.sse. asses G SYSTEM ENGINEERING /

INTEGRATION I e REACTOR SYSTEM S REACTOR VESSEL AND INTERNALS G BUILDING AND STRUCTURES p

1

' e HEAT EXCHANGERS S POWER CONVERSION SYSTEMS e SHUTDOWN HEAT REMOVAL 8 CONTROL ROOM DESIGN SYSTEM S AUXILIARY SYSTEMS

! # LICENSING SUPPORT S REFUELING SYSTEMS e LICENSING SUPPORT e SAFETY AND LICENSING i

i I S MARKET / COMMERCIALIZATION S MARKET / COMMERCIALIZATION ASSESSMENT G MARKET / COMMERCIALIZATION I o ANL ETEC HEDL LANL ORNL Al WAESD e PREHEAT DESIGN e FUEL CYCLE oCORE o SHIELDING e STEAM e PIPING e CONTROLS / e CORE DATA ENERGETICS ANALYSES ERATOR SUPPORT e CORE ANALYSIS PORT

• YS e A ALS/ STRESS e SitR TIER AL/

• 1RANS4ENT e PLANT EXPERIENCE p y

'EE' "; :== =A e = =E

• SAFETY ANALYSES e STRUCTURES

=r e NA AEROSOL ANALYSES

- = r-

• SASS SUPPORT

' e NONNUCLEAR SUPPORT

,. .pgAsuppOHI I&C ANALYSIS i e THERMAL / e MMI e HUMAN FACTORS N HYDRAULIC SUPPORT T ANALYSIS

@ . SwReS ANatvSiS

.I -

i g e _R.c . . - , -_ -

_ . - ggly Ts>.s.

l lI . _ _ _

l 8

SAFR PROGRAM LOGIC l THE INITIAL SAFR DESIGN CONCEPT IS BEING DEVELOPED AND IMPROVED OVER THE COURSE OF THE 37-MONTH PROGRAM, BASED ON TEST AND ANALYTICAL 5 RESULTS OBTAINED FROM THE BASE LMR TECHNOLOGY PROGRAM, USER l

NEEDS AS INDICATED BY UTILITY INTERACTIONS, FEEDBACK FROM NRC DIS-CUSSIONS, AND GUIDANCE RECEIVED FROM A" BLUE RIBBON" ADVISORY COM-l' MITTEE. THE LATTER COMMITTEE WAS FORMED TO AlD THE TEAM IN ADDRESS-ING THE VARIOUS !NSTITUTIONAL ISSUES CONFRONTING THE PROGRAM. THIS g

~

PROGRAM WILL RESULT IN A STANDARD DESIGN TAILORED TO MEET USER'S

] NEEDS AND IN PLANS AND COST / SCHEDULE ESTIMATES WHICH WILL PROVIDE A l 1 BASIS FOR PROCEEDING WITH THE EVENTUAL COMMERCI AL INTRODUCTION OF THE CONCEPT.

i l

l 5

l3

• j i

i $

m nock am f sean m ) - - 15-39-0 g

I i

1 SAFR PROGRAM LOGIC lB l

i h _.

3 INSTITUTIONAL I 'i I

ADVISORY ' ,

/C COMMITTEE - l- '

%k .s[t kk)

/ /

I l

j DESIGN EVALUATION / PLANNING FOCUSED ON USER NEEDS y,.,

f. '.-}'I ._ h f-

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DEM T A ION

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a g v o

1. DESIGN MEETING USER NEEDS SAFR s LICENSING PLAN '

STA '"

pOIN h '

iEEII.l_ 25F 5 NRC LICENSING POSITION DESIGN im 11_t1 * $1 COMMITMENT I 8 ORGANIZATION /

l MANAGEMENT PLAN I

BASE UTILITY NRC PROGRAM INTERACTIONS DIALOG

• FINANCIAL PLAN 4 R&D # SUPPLY PLAN y e POTENTIAL l g CUSTOMERS l l 5 =

l E!

l i

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CCL L-v L-099 j us uma num mes. - _ _.

I

) O O O i l I l 8 TRANSLATING MARKET NEEDS INTO SAFR GOALS l 3 THE GOALS SUMMARIZED ON THE NEXT TWO PAGES WERE ESTABLISHED FOR g SAFR AFTER ANALYZING CAPACITY PROJECTIONS AND THE COMPLEX OF CON- E l CERNS WHICH TODAY CONFRONT THE UTILITY INDUSTRY. ACHIEVEMENT OF

) THESE GOALS IN A PLANT DESIGN SHOULD GO FAR TOWARDS ALLEVIATING g l

UTILITY CONCERNS AND PROVIDING A VALUABLE ELECTRICAL POWER OPTION.

i

! I i

1 I

I I

=

e I

5:

l Rockwellintemational seassusteen)esessmannens 15-40-0

. ..... ~ -

4

l O O O i

SAFR GOALS

I

' g e ASSURED PUBLIC PROTECTION e ESTABLISH SAFE CONDITIONS WITH NATURAL FORCES e AVOID NEED FOR FAST OPERATOR l ACTION g e OBVIATE NEED FOR EVACUATION i

e ASSURED INVESTMENT PROTECTION -

l I e LOW PLANT DAMAGE PROBABILITY e EXCEED ASME-SC-C <10-*/ YEAR e LOW ACCIDENT-CAUSED e EXCEED ASME-SC-B <10 / YEAR

! DOWNTIME

I e COMPETITIVE, CREDIBLE ENERGY e <50 mills /kWh t COSTS i e SHORT ONSITE CONSTRUCTION e <4 YEARS "

! SCHEDULE l

1. HIGH CAPACITY FACTOR e >80%

l c,

• LOW PERSONNEL RADIATION e <25 MAN-R/ YEAR -

l g 3 EXPOSURE i

l $

e e.=,_- @) -->-- 16-19-0

]

!I

4 i I SAFR GOALS l

I

g e WIDELY USABLE STANDARD PLANT e >75% OF U.S. SITES

• LIMITED FINANCIAL RISK e MINIMlZE TIME TO INCLUSION OF l l INVESTMENT IN RATE BASE e MINIMlZE UNCERTAINTY IN PREDIC-l!li TION OF CAPACITY REQUIREMENTS /

PLANT COST, SCHEDULE jll e CLOSELY MATCH CAPACITY ADDI-l TIONS TO GROWTH DEMANDS l e. MINIMUM DEVELOPMENT RISK AND e USE EXISTING TECHNOLOGY; DESIGN COST COST-EFFECTIVE TEST PROGRAM

~

l e MINIMlZE COMPLEXITY OF MANAGING e SIMPLIFY DESIGN

• MINIMlZE ONSITE LICENSE-RELATED h WORK 4

j

• LONG PLANT LIFE * >60 YEARS g

! a o EFFECTIVE SECURITY AND SAFE- 0 CONSIDER IN DESIGN FROM OUTSET g GUARDS MEASURES I a l =

l e "*"~~ ~~.._-gpy _ >.._.._. . ,o.e .

l E

I -- _ - - - - -

O O O w MAJOR FEATURE SELECTIONS l THE MAJOR FEATURES OF SAFR WERE CHOSEN AFTER THE TEAM, IN CLOSE E W

COOPERATION WITH ARGONNE NATIONAL LABORATORY AND CHICAGO BRIDGE l AND IRON, CONDUCTED A COMPREHENSIVE, DETAILED REEVALUATION OF THE j OPTIONS AVAILABLE. THIS CHART INDICATES VARIOUS OPTIONS CONSIDERED l FOR SOME OF THE MAJOR FEATURES. IN PARTICULAR, THE 350-MWe MODULE SIZE, CHOSEN AFTER A LENGTHY AND DETAILED STUDY, INDICATED THAT THIS g

, SIZE OFFERS SIGNIFICANT COST ADVANTAGES OVER SHIPPABLE SMALLER a UNITS, AS WELL AS AN OPTIMUM BALANCE OF FACTORS, PROVIDING I

e COST, SCHEDULE, QUALITY ADVANTAGES OF SHOP FABRICATION e ECONOMY OF SCALE E e MINIMUM FIELD CONSTRUCTION AND TIME TO STARTUP '

5 e DIVERSE, PASSIVE SHUTDOWN HEAT REMOVAL SYSTEMS I

I I

=

e I

e=== g .. _ _ >. _ . _ _ I l l

i O O O i l I

l MAJOR FEATURE SELECTIONS l 5 SAFR l CHARACTERISTICS v

! FEATURE OPTIONS EVALUATED BEST CHOICE l 8 PLANT SIZE / e 1000-1350 MWe MONOLITHIC (FIELD -

G 350 MWe BARGE SHIPPABLE i ARRANGEMENT FABRICATION) MODULAR PLANT -SHARED i e FACTORY FABRICATED / SHIPPABLE 110-350 MWe FACILITIES j MODULES e NSSS e POOL, LOOP e POOL i CONFIGURATION e FUELTYPE -

e OXIDE, METAL e INHERENTLY SAFE METAL

CORE .

1

{ e STEAM CYCLE e SATURATED STEAM e 950*F REACTOR OUTLET,

4 SUPERHEATED STEAM -- SULZER CYCLE SUPERHEATED BENSON e SUPERHEATED STEAM - BENSON CYCLE STEAM CYCLE

]

) e SAFETY SYSTEMS e REACTOR SHUTDOWN - DIVERSE, REDUNDANT, e REDUNDANT,61 VERSE PAS-

! SELF-ACTUATING SIVE SAFETY SYSTEMS

! e SHUTDOWN COOLING - FORCED AND NATURAL CONVECTION, VARIOUS LOCATIONS IN HEAT l TRANSPORT PATH i <c 1

e RocketoyneDvisk>n gy -->--. 17-l1-l

, I _ - _ _ _ .

O O O SAFR UNIT FLOW DIAGRAM I

3 THIS SHOWS THE FLOW DIAGRAM FOR THE 350-MWe SAFR STAND- g ARD UNIT. THIS DESIGN IS BEING DEVELOPED WITH DESIGN EMPHA-SIS ON (1) SELF-PROTECTIVE SAFETY FEATORES, (2) SIMPLICITY AND g

LOW CONSTRUCTION QUANTITIES, (3) MINIMlZING AND LOCAllZING j THE SYSTEMS / COMPONENTS UPON WHICH NUCLEAR SAFETY DEPENDS, AND (4)" USER FRIENDLINESS."THE DESIGN IS BASED g

ENTIRELY ON EXISTING TECHNOLOGY. THE REACTOR ASSEMBLY IS g

SHOP-FABRICATED AND SHIPPED TO THE SITE. ESSENTIALLY ALL .

RADIOACTIVE MATERIAL IN THE PLANT IS CONTAINED WITHIN THE g

40-ft-DIAMETER POOL-TYPE REACTOR VESSEL. THE DIAGRAM ALSO SHOWS FEATURES INVOLVED IN THE DIVERSE, PASSIVE DECAY HEAT REMOVAL SYSTEMS, AS WELL AS THE CONVENTIONAL ONCE g

THROUGH SUPERHEATED STEAM CYCLE EQUIPMENT. UNITS SUCH g

AS THIS CAN BE JOINED IN ARRAYS AS NEEDED TO PROVIDE THE

$ REQUIRED GENERATING CAPACITY.

• E

@ u e I

@ =::==-' di$ -->-- u ,o.o I I

i I O O O l I 350 MWe STANDARD SAFR UNIT FLOW DIAGRAM l

I I G 350 MWe I _

PUMP j 2 00 psig 620*F g p

• l NET r )

l AIR Jgoo % g k Q  !. l l . ,,-...v..

lHXs  :

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% y m l 1 s i . EE y- >

l

, f gg LO S -- ,

950*l- 4 l l .

i E

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A

. CONTAINMENT BOUNDARY g ..

+  % . r{ +

r CORE -h W T< m I */%

I 1

4 1 I .

L^ 8^FETY -= = HIGH QUALITY INDUSTRIAL STANDARDS GRADE STAND RD g e .m . .. _ -

g .. _ _ _ >.. _ ... _

1 i I - _ - _ _ _ _ _ - _ _ -

l SAFR NUCLEAR ISLAND I E

I THE FIGURE OPPOSITE SHOWS A CUTAWAY VIEW OF THE REACTOR ASSEMBLY I

WITHIN THE RECTANGULAR, REINFORCED-CONCRETE REACTOR BUILDING. SAFR l EMPLOYS A POOL-TYPE REACTOR IN WHICH THE PRIMARY INDUCER PUMPS AND INTERMEDIATE HEAT EXCHANGER ARE POSITIONED WITHIN THE REACTOR VESSEL AROUND THE CORE. ALL PRIMARY RADIOACTIVE SODIUM IS CONTAINED WITHIN THE l

i SIMPLE CYLINDRICAL REACTOR VESSEL, WHICH HAS NO NOZZLES, ATTACHMENTS, l

OR PENETRATIONS.THE REACTOR INTERNALS AND DECK STRUCTURE ARE l DESIGNED SUCH THAT THE COMPONENT PARTS OF THE ENTIRE REACTOR ASSEM- l BLY CAN BE INTEGRATED WITHOUT BIMETALLIC JOINTS. /

HATCHES IN THE REACTOR BUILDING ROOF PERMIT VERTICAL HANDLING OF EQUIPMENT ITEMS, THUS MINIMlZING BUILDING HEIGHT. SPENT FUEL IS DRAWN g

FROM IN-VESSEL STORAGE BY THE FIXED A-FRAME FUEL HANDLING SYSTEM, AFTER WHICH IT IS EITHER SHIPPED OR REPROCESSED ON SITE. THE NON-SAFETY-GRADE g INTERMEDIATE SYSTEM AND STEAM GENERATORS ARE HOUSED IN CONVENTIONAL l INDUSTRI,AL BUILDINGS. g i

i l

U ,$ ,*,", U ,5 7 17-3-1 l

l l

i O O O l Sodium ADVANCED g. td I .

[ '

~ '

FAST '4 k-l QEACTOR ,

4

[

!1' j

,, {

l NUCLEAR ISLAND g e%

<g -. );

l .

I , -

~

,6 u g J'>- 'l l \ ()!.-

l '

l, d ,

l .

lv z j, ..

i I ( .

l e =_-

, 2

. u m a .....

i O O O i I

4 SAFR UNIT ARRAY - 1400 MWe I

i l

THIS ARTIST'S CONCEPT ILLUSTRATES THE USE OF MUL-j TIPLE STANDARD 350-MWe UNITS - IN THIS CASE FOUR g

" MODULES" WHICH COMBINE TO PROVIDE 1400-MWe E l GENERATING CAPACITY. 350-MWe MODULE SIZE WAS FOUND TO BE THE OPTIMUM TRADEOFF OF SHOP-

! FABRICABILITY, LOW CAPITAL COST, PASSIVE DECAY HEAT

! REMOVAL SYSTEMS, SHORT CONSTRUCTION SCHEDULE, ,

i AND PRODUCT INVESTMENT RISK. THE TURBINE GENERA-TOR FACILITIES ARE SEPARATE FROM THE NUCLEAR

! ISLANDS AND CAN BE CONSTRUCTED USING HIGH-l QUALITY INDUSTRIAL STANDARDS WITH CONCOMITANT

! COST SAVINGS AS COMPARED WITH NUCLEAR SAFETY- -

RELATED STANDARDS.

I a i  ! E i  ;

i i I

@ =:==- @ -->-- iz.e.,

3

I,

I O -

- z . ., - .

O . ~ . x O

I .,

-.---.M, a

t3-n;pl' .,#* * ; 3 ,-

h -w,wu .

, , , , ;g q -

a -rM&Lu::q $.w . . m .f 3 ,:

. , Q '

g

_,00!4 SAFR UNIT ARRAY 1400 MWe "o

c. p:":G;4?.;gme.g3ygr .f :;. p v = --

I .

l

' ~

,+

s d CONTROL

+. .

, g 'f * * ,,,.h' . . . _ . BUILDING

• :' i

.f.

':pj )hl .a .

FUEL CYCLE f.:j;Q:., '

FACILITY

!A '

-v,y.

I , ..

=:..

. ..~. ,

,a -s ' . % ,

~

x %Qs .

, , . . .  %. c y

I TURBINE GENERATOR UILDING .

saf,. 4

'5 '

NUCLEAR ISLAND MAINTENANCE

/ .+r- -

BUILDING NUCLEAR FUEL ISLAND

' TRANSPORTER - -

E t l @ =._ . . . --

I

E -

- - - . ~ ~

g 2.. :1 s::. v. .

. a w ; w . a ,. :.,... .3 ; , . , 3 y

.'.. . o : n aa.. ....

1400 MWe ~ ~ " -

.- ..v:~,.t,y,4 4 SAFR UNIT ARRAY I

' ~

I.b #$'$i,h e Kriay & Q.~*..A p j' . , ;/"7'C~7" s

, _.. .},y I

I .:.+

.a 4 .: CONTROL ht*j.

."' BUILDING t[ N.I ' ,gggif ~

.I.i FACILITY p;.);ji[T .

I.'l'i . .+ p, ..,

5 p

g k& ,

'x A ., N . ,

h - 6 %. ,

• I -

9 TURBINE '5 '

NUCLEAR GENERATOR y,.. v. ,

BUILDING . . *: "' ISLAND MAINTENANCE

, 2.c ..

BUILDING NUCLEAR 1-

.. . s

, FUEL ISLAND '

- '~ TRANSPORTER i e- _

a

i SAFR CONSTRUCTION COMMODITIES MINIMlZED l

I I

l<

I 1

g

! A MAJOR GOAL IN THE SAFR DESIGN WAS TO MINIMlZE PLANT CON-

STRUCTION QUANTITIES, CONSISTENT WITH PROVIDING THE l

1 ACCESS AND SPACE NECESSARY FOR EFFICIENCY IN PERFORMING l

PLANT MAINTENANCE AND INSPECTION FUNCTIONS. LOWERING CONSTRUCTION QUANTITIES REDUCES COSTS AND SHORTENS

?

l SCHEDULE.THE SAFR CONSTRUCTION COMMODITIES ARE SUB- l STANTIALLY LOWER THAN CORRESPONDING LWR FIGURES, AS i

ILLUSTRATED ON THE PAGE OPPOSITE. .

l l

I is . -

I' l

b j

bhWen worsaw Rocketdyne Divisaon [ s-sim)e-a- 3743 g jl I

O O O

~

I I COMPARISON OF TOTAL PLANT g

CONSTRUCTION COMMODITIES I

I PWR*

1139 MWe SAFR 4 x 350 MWe l CONCRETE

[ CUBIC YARDS (yd/MWe)] 174,000 (153) 160,000 (114)' '

5 REBAR [ TONS (T/MWe)] 28,000 (25) 14,000(10) l STRUCTUR AL STEEL

[ TONS (T/MWe)] 12,000 (10) 9,300 (7)

LINER PLATE

[f12 (ft2/MWe)] 160,000 (123) 100,000(71) '

I CABLE [1000 LINEAR FEET (1000 ft/MWe)] 7,400 (6) 5,800 (4)

REFERENCE:

UE&C-ANL-82093 -

O I

O e, :,==- @f -->-- is.2i-i

.I

!I

O O -

O i SAFR SIMPLIFIES MANAGEMENT TASK

- I I

I I

THE SAFR APPROACH SHIFTS THE FOCUS OF NUCLEAR SAFETY-RELATED FABRICATION ACTIVITIES FROM THE FIELD TO THE SHOP, WHERE APPROXIMATELY 70% OF THIS WORK IS PERFORMED.THIS ENHANCES PRODUCT QUAL-ITY, REDUCES LABOR COSTS, SHORTENS OVERALL CON-STRUCTION SCHEDULES, AND SIMPLIFIES THE UTILITY AND NRC TASKS ON THE PROJECT.

I

?, .

I e ====- @ -->-- ,m 3

5.

E SAFR SIMPLIFIES MANAGEMENT TASK I .

I MONTHS -30 -2'O -1'O 2'O 3'O 1[0 4'O g START I FACTORY SITE WORK I INUCLEAR SAFETY-RELATED l

REACTOR ASSEMBLY I {

^

HEAT TRANSFER EQUIPMENT FUEL HANDLING I TURBINE GENERATOR FIELD CONTAINMENT BUILDING I CONTROL BUILDING M

l SERVICES BUILDINGS TG BUILDING POWER I m STARTUP/ TEST I

OPERATION U

I b

I e '

g q p_ ,.._.,.,,

,_,- g . ..

I I _ - _-

l CT O O g SAFR FUNDAMENTAL FEATURES WHICH ENHANCE AVAILABLILITY l - AS COMPARED WITH LWRs -

I e AUTOMATED UNDERHEAD REFUELING WITH PERMANENTLY INSTALLED l EQUIPMENT

• FAR FEWER ENGINEERED SAFETY FEATURES- '

' l

• FAR FEWER AUXILIARY SYSTEMS l

• NONRADIOACTIVE TURBINE GENERATOR, STEAM GENERATOR, AND WATER j SYSTEM MAINTENANCE / INSPECTION OPERATIONS

l

• FEWER PPS SIGNALS / FEWER SPURIOUS SCRAMS ,

I g

• NO BURNOUT-DICTATED TASK INTERRUPTION / PERSONNEL SHUFFLING t E e LMR OCCUPATIONAL EXPOSURES ~1/10 LWRs

• MINIMAL RADWASTE HANDLING AT SITE ,

• NO SPRING HANGERS / SNUBBERS IN MAIN SODIUM SYSTEMS, DRACS l e NO HIGH PRESSURE SEALS (>150 psig) IN RADIOACTIVE SYSTEMS E e MAINTENANCE / INSPECTION PREPARATIONS POSSIBLE DURING OPERATION l IN E'SSENTIALLY ALL AREAS s

( nockweii non.i . ..n ). .

38 22-0 I .

O O O I-LMR CONVERTER DESIGN IS ALMOST IDENTICAL TO BREEDER l ,

I

~

l I;

COMMERCIAL DEPLOYMENT OF ADVANCED LMRs WILL ENABLE THESE PLANTS TO MAKE A CONTRIBUTION l TOWARD MEETING THE NATION'S ENERGY NEEDS OVER THE INTERMEDIATE TERM. AT THE SAME TIME, THEY WILL LAY A SOLID GROUNDWORK FOR A FUTURE BREEDER

• li ECONOMY. AS MAY BE SEEN FROM THE TABLE OPPOSITE, LMR CONVERTERS ARE VIRTUALLY IDENTICAL IN DESIGN l

TO BREEDERS EXCEPT FOR A FEW DIFFERENCES IN CORE l DESIGN AND REFUELING PROVISIONS.

I e I O =:::.':=#-' @) -->-- u.e.o l I

i O O O I LMR CONVERTER DESIGN IS ALMOST lDENTICAL TO BREEDER l

I SAFR UNIT 350 MWe BREEDER CORE DESIGN

- INCREASE NUMBER BLANKET ELEMENTS /

, g PIN DIAMETER 5

REFUELING

- INCREASE STORAGE g POSITIONS ~13 SAME l OVERALL REACTOR ASSEMBLY SAME -

PHTS I IHTS SAME g BOP SAME FUEL CYCLE SAME OPERATIONS SAME g SAFETV/LICENSABILITY ~SAME l a, y e =:==- @) -->- is.33.o I

I O O O I

I i I I

I 4

I

SAFETY GOALS / FEATURES I .

I n

I $ '

a I

O =::==- di8) -->--

I I - - - - -

l O O O u SAFR TOP-LEVEL SAFETY GOALS I

l SAFETY HAS FIRST PRIORITY IN THE SAFR DESIGN. THE TOP-LEVEL GOALS ESTAB-l LISHED FOR SAFR DESIGN GO BEYOND THE EXISTING NRC/ EPA DOSE AND RISK l CRITERI A. THE DESIGN INCORPORATES INHERENT SELF-PROTECTIVE FEATURES WHICH ENABLE IT TO MEET THE GOALS PRESENTED ON THE FACING PAGE NEITHER ACTIVE SYSTEMS NOR OPERATOR ACTIONS WILL BE REQUIRED TO PRO-TECT THE PUBLIC AND THE UTILITY'S INVESTMENT FROM THE EFFECTS OF DESIGN l

BASIS ACCIDENTS. SAFR PROVIDES THIS ULTIMATE NATURAL SAFETY IN ADDITION TO THE PLANT PROTECTION SYSTEMS WHICH PROVIDE RELIABLE REACTOR PRO-l TECTION UNDER NORMAL CIRCUMSTANCES. REDUNDANT AND INDEPENDENT SHUT-DOWN METHODS ASSURE TERMINATION OF REACTOR TRANSIENTS, AND CORE f l

COOLING AFTER SHUTDOWN IS PROVIDED BY PASSIVE SYSTEMS WHICH DO NOT DEPEND UPON THE ACTION OF ANY VALVES, SWITCHES, OR ACTUATORS.

l BY EMPLOYING LAYER UPON LAYER OF PASSIVE, INHERENT SAFETY FEATURES, DESIGN LICENSABILITY IS ENHANCED AND IT IS POSSIBLE TO GREATLY SIMPLIFY g AND REDUCE THE NUMBER OF ACTIVE SAFETY SYSTEMS, SAVING MILLIONS OF DOL-g LARS IN PLANT COST. ELIMINATING THESE SYSTEMS ALSO INCREASES PLANT g y AVAILABILITY BY ELIMINATING NUMEROUS SOURCES OF COMPONENT FAILURES WHICH COULD CAUSE PLANT SHUTDOWN. g

~

e ==,e @ -->-- ,,.,., g g

I O O O I SAFR TOP-LEVEL SAFETY GOALS (PROVIDE SAFETY EQUIVALENT TO LWRs BUILT IN SAME PERIOD)

I

# INHERENT SAFE RESPONSE TO ALL CREDIBLE EVENTS, INCLUDING e TRANSIENTS WITHOUT SCRAM l

e LOSS OF NORMAL DECAY HEAT REMOVAL e STRUCTURAL FAILURES I

1. MINIMlZE POTENTIAL FOR SEVERE ACCIDENTS e INCORPORATE RELIABLE, DIVERSE SHUTDOWN SYSTEMS g e LIMIT CHALLENGES TO PLANT TRIP SYSTEM e MINIMlZE DEPENDENCY ON SUPPORTING SAFETY EQUIPMENT .

• PROVIDE LONG TIME PERIODS FOR CORRECTIVE ACTION

• ELIMINATE NEED FOR EVACUATION g e PROVIDE DEFENSE IN-DEPTH WITH MULTIPLE BARRIERS -

e EMPLOY LOW LEAKAGE CONTAINMENT l e ASSURE LOW CORE MELT PROBABILITY e DESIGN FOR LOW PRA RESULTS e

I 5 .

z I

• I O ==-- @) - ->- - ,4.es.o I - _ - _ _ -

O O O i ASSURED SAFETY OF PUBLIC, INVESTMENT I The SAFR plant takes full advantage of the favorable safety-related properties of the sodium coolant (e.g., high boiling point, low vapor pressure, excellent thermal conductiv-l ity, strong natural convection) in providing safety assurance. Some of the important plant g safety features are summarized below and highlighted on the chart opposite. E REACTOR SHUTDOWN e Redundant / diverse control rod systems, including self- E actuated secondary system 3 e inherent thermal and nuclear feedback shuts down after loss of flow without scram g DECAY HEAT REMOVAL e Natural convection air cooling of vessel e Na-Na-air natural convection direct reactor auxiliary t l

cooling system g STRUCTURAL COOLING e Natural convection air cooling of deck and cavity 'l.

i STRUCTURAL INTEGRITY -

e Redundant box beam deck structure g

e Compressive core support system a

y GENERAL M

w e Low core energet.ics 3

$ e Long grace period (>24 h) for operator action E h Q*j','"l"*" "*' seasmusnon) msmaamma 2-12-1 j' i

i O O O I SELF-PROTECTIVE REACTOR FEATURES .

ASSURED DECAY HEAT REMOVAL e NATURAL CONVECTION MR I DIRECT REACTOR AUXILIARY COOLING SYSTEM (DRACS) e NATURAL CONVECTION f

ASSURED REACTOR SHUTDOWN l

lEP YSTEM RA S , d-

\ *** *

• O REDUNDANT, DIVERSE g REACTOR SHUTDOWN

-h .

G INHERENT, SELF-ACTUATED SHUTDOWN

'3 dg- . INHERENT, SELF-LIMITING FEEDDACKS y !g . - A '

STRUCTURAL REDUNDANCY USER FRIENDLY j# -

e CONTAINMENT VESSEL I e ALL RADIOACTIVE MATERIAL IN SIMPLE ENCLOSURE *,

. I

( _

o

[ 'j)d-M ASSURES ADEQUATE COOLANT INVENTORY O PASSIVE, COOLED DECK '

e THERMAL INERTIA p* ._

ALLOWS LONG PERIODS iJ -

9 COMPRESSIVE CORE SUPPORT FOR OPERATOR ACTION--

\

I *;

4

?

-SODIUM FIRE RISK AVOIDED 4 INERT ATMOSPHERE (h *. [. , e GUARD VESSEL /PlPES I h" 4

//YM' [* , . *

..s...- *

[ %' e IN STEAM GENERATOR BUILDING - TESTED FIRE SUPPRESSION FEATURES B-E m _ Il m, _ . -

g Rocketdyne g-Division . ..

I i

I ___ __ _ _ _

O O O i LMR SELF-REGULATING FEATURES '

I I

~

I LMRs CAN, BY CAPITALIZING ON SODIUM PROPERTIES, BE I

DESIGNED TO BE SELF-REGULATING / POWER LIMITING j UNDER VERY SEVERE CONDITIONS. THIS EFFECT IS ILLUS-TRATED BY THE EXPERIMENTAL RESULTS SHOWN ON THE j OPPOSITE PAGE FROM THE EBR-il AND FRENCH RAPSODIE -

REACTOR. SAFR INCORPORATES THIS CAPABILITY, AS ILLUSTRATED BY THE REACTOR RESPONSE TO VARIOUS l

ATWS EVENTS AS SHOWN ON THE FOLLOWING PAGE j

~

I 1 s. .

1 e gg.3,=y g . ..

,,.,., y

, I,

s I

g LMR SELF-REGULATING FEATURES

., EBR-il - LOHSWS t e mir se SAFRLOFWS

- 1

/ vaurenAtuns E , , , ,

f a _

PEAuAssameLv I 0, im y

es a p ises t j goes _'*76TsA Un1 to#s _

~

i- J Po..n -

se e_ ,,

n, i

N nan COOLANT y *u see ou m m aos m ese ne Rim - -

~ c"'f."_' "s t---

• r- -~ 80s I

T EBR-li - LOFWS e aos ses ***

inse t.:

I SAFR TOPWS F, se _

I

~

emir TEuPinAlunE R asse

~~

c66iist iAruniisoN

~ ~ *

"g a _ see.

se = F

$ 800 2

h" L(

I 2im i ,,, f rowen

~

8' E

g ease _

PtAu coot ANT

- sc

'~is%e aos $e-$0 Ne

• Iso E- cool ANT INLET

~~~ "'~"~~

I Tinse fit *

,_ p .,

RAPSODIE - LOFWS e * * * *

-, .--r as T'"8 8'l

'shdidss6iiissP6sii"'~ SAFR LOHSWS .

p" sees- c'ooLANT s"ArunAr[oM h {fy,'n',^r*uN '8i -

E a

e. i.es-I E

=

i. i 5= M' 2 im

{

y.-

1. " E",' ~~

POwtn

, [ "*" c99M'd-h g

,,,e ese ,e...~~,

cooLANi mar 4 &

r e e - de ees - de- - a ,eee . de se et,e see

= i>

I -

y g . ...... .. _ -

O..,.. gy..___3.._..._

O O O i PASSIVE, DIVERSE NATURAL CONVECTION DECAY HEAT REMOVAL SYSTEMS 3 I

THE THREE DIVERSE, INDEPENDENT METHODS OF DECAY HEAT REMOVAL - NAMELY, THE NORMAL HEAT TRANSFER PATH, THE DIRECT REACTOR AUXILIARY COOLING SYSTEM (DRACS), AND THE REACTOR AUXILIARY COOLING SYSTEM (RACS)- g ASSURE THAT THE REACTOR WILL NOT EXPERIENCE EXCESSIVE TEMPERATURES. m THE FACING CHART SHOWS THE NATURAL CONVECTION RACS AND DRACS COOLING CIRCUITS.THE RACS IS IN OPERATION AT ALL TIMES, ASSURING ITS CONTINUAL READINESS FOR PERFORMING THE DECAY HEAT REMOVAL FUNCTION.THERE IS A 0.3% HEAT LOSS THROUGH THE RACS SYSTEM UNDER NORMAL POWER OPERATIONS.

THE DRACS NATURAL CONVECTION FLOW INITIATES AFTER REACTOR SHUTDOWN.

THE CHART ON THE FOLLOWING PAGE SHOWS THE A-FRAME-TYPE FUEL HANDLING .

I SYSTEM. SPENT FUEL IS STORED WITHIN THE VESSEL UNDER SODIUM FOR.1 YEAR j PRIOR TO TRANSFER TO EX-VESSEL WHERE IT IS COOLED BY THE NATURAL CON-VECTION OF AIR. THE DECAY HEAT LEVELS AT THE END OF THE IN-VESSEL STORAGE j PERIOD ARE SUFFICIENTLY LOW THAT ALL ASSEMBLIES CAN BE COOLED ADE .

$ QUATELY BY NATURAL CONVECTION IN THE EVENT THEY BECOME STUCK AT ANY j

• POINT ALONG THE FUEL HANDLING PATH.

~

I e ===- @) -->-- m e.o i I

O O O l PASSIVE, DIVERSE NATURAL CONVECTION DECAY HEAT REMOVAL SYSTEMS I

.f Enri.. 9gg=

E

%r

/ W~ - - O

.yg r

J:Y,

~

l .

i "$Ifv"$'now g '

2 3

Onn'c'a^a .

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l

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w 1

, 2)'  ::

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~;

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(if7" rl l 'I l k,( - [,[a,1,$ a'uo.

L it i; g ~

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jD bea lh j l l

g

- D l:..: .

[ . .

g $ aoa ._ _w w @>

l I

i l PASSIVE SPENT FUEL COOLING Iht NATURAL CONVECTION GAS COOLING h J OF ALL ASSEMBLIES AFTER POWER - KW INVESSEL STORAGE /A /

MINIMUM 16.8 3.1 l

g MAXIMUM 25.4 12.4 l

g TOTAL 551 150

! EXVESSEL STORAGE .

M XI UM 8 .9 _ '. ~ "

~g ~ ~ ~ '

- - ~---J

~ ~ ~ ~

TOTAL 40 11 ll I t

= j'm ii l,'

A

[

3 INVESSEL STORAGE

,, /_ <:

j g C ,j/ s R

{[L_

/ _-

l l i l$7

E '

PEAK ASSEMBLY TEMPERATURES (*F)  ! ,

~

F/A STUCK IN REACTOR PORT 463 J ~

' d ij  ;

j F/A STUCK IN FTC 319 / '

j F/A IN INTERIM STORAGE 450 .- c - -

h 9  !<

l $ FUEL STORAGE E -

COOLING AIR IN 3 o j y > = =- ggty .._._3._.._

I ~

E

O O CF I

I DECAY HEAT ACCOMMODATED WITH MARGIN THIS PLOT OF HOT POOL TEMPERATURES FOLLOWING PLANT SHUTDOWN g SHOWS THE ABILITY OF SAFR TO ASSURE ADEQUATE CORE COOLING, EVEN WHEN THE LOSE OF NORMAL AND DRACS SYSTEMS IS POSTULATED. IN FACT, IF E

THE LOSS OF ALL COOLING IS POSTULATED, THE LARGE THERMAL INERTIA OF 3 SAFR RESULTS IN A "LONG" GRACE PERIOD IN WHICH TO TAKE CORRECTIVE ACTION. j I

R i

g 3 e

i I O =::.:=- @) -->--

g 15-45-0 1

s O O O I DECAY HEAT ACCOMMODATED WITH MARGIN I

i i I I I I I E

I 1600 -

~ ~ ~ ~ ~ ~

I 5

=

1 1400 I

3 1350 MAJOR PLANT l-DAMAGE (FUEL)

}5 1200 250 SHUTDOWN FOR I -

INSPECTION O --

M

" p ~~ RACS (10-3/ YEAR) '

I O a 1000

-[ 'N NO PLANT DAMAGE O /

r DRACS + RACS (0.2/ YEAR)% %

800 I I I I I I I 0 10 20 30 40 m TIME AFTER REACTOR SHUTDOWN (h) ,

t !s -

I e e .m_,, ._,-

gy.___>...._

E _ - _ - -

O O O i I

DIVERSE SELF-ACTUATED SHUTDOWN SYSTEM j THE PRIMARY SHUTDOWN SYSTEM FOR SAFR IS THE WELL-TESTED CRBR SYS-TEM.THE DIVERSE, SECONDARY SYSTEM IS A SELF-ACTUATED CONCEPT WHICH HAS BEEN UNDER DEVELOPMENT FOR SOME TIME ON DOE's LMR BASE PRO- '

GRAM. IN THIS CONCEPT, ARTICULATED ABSORBER ELEMENTS ARE SUPPORTED g ABOVE THE CORE BY A MAGNETIC LATCH.THE LATCH IS AN INTEGRATED CON-FIGURATION SUCH THAT INCREASES IN SODIUM EXIT TEMPERATURE FROM THE CORE ARE RAPIDLY COMMUNICATED TO THE LATCH. IF THE OVERTEMPERA-TURE IS SUFFICIENT, THE CURIE POINT OF THE LATCH MATERIAL IS REACHED AND THE ABSORBER IS RELEASED - DROPPING INTO THE CORE AND SHUTTING g DOWN THE REACTOR. THE PLOT OF TEST DATA ON THE OPPOSITE CHART SHOWS THAT THIS DEVICE PROVIDES A GOOD MARGIN (EVEN UNDER WORST CASE TOLERANCE STACKUPS) FOR SAFE SHUTDOWN OF THE REACTOR.

h I

a .

31

$ I Oms.=- di) -->-- , , . . , 3 1

O O O I INDEPENDENT, DIVERSE, REDUNDANT 5 SHUTDOWN SYSTEMS E

LOY PRIMARY SECONDARY ~/ /{ N ACTUATION * " '

APTS (2-OUT-OF-4) 1. MAGNETIC FLUX SWITCH / Na = p'~/ T4TCH I

OVERTEMPERATURE 2.PPS l .,

N ARTICULATED RELEASE MECHANISM ROLLER NUT MAGNETIC LATCH $8'SEG TS)

INSERTION SPRING ASSISTED GRAVITY GRAVITY CONTROL ASSEMBLY SINGLE RIGID ABSORBER ARTICULATED ABSORBER ,

I PERFORMANCE ACHIEVES COLD SHUTDOWN ACHIEVES HOT SHUTDOWN WITH SINGLE FAILURE

• WITH SINGLE FAILURE * ,

/ DASHPOT PROVEN DESIGN CRBRP, FFTF, LSPB LMFBR BASE PROGRAM - SASS y

= ll a INHERENCY FAIL-SAFE ON LOSS ACTUATES ON LOF, TOP '

OF SERVICES 10I 10I 101

• lNCLUDING STUCK ROD - U U U--

y 1 5 a

p

~ 3 .m ._

- g . ..

E

r O O O g

FAR FEWER ENGINEERED SAFETY FEATURES REQUIRED I

I O TYPICAL LWR SAFR S DECAY HEAT REMOVAL RHRS (ELECTRICAL POWER RACS AND DRACS (NO ELEC-SYSTEMS REQUIRED) TRICAL POWER REQUIRED)

AUXlLIARY FEEDWATER SYSTEM -

• G SAFETY INJECTION SYSTEMS HPSI, LPSI, BORON SYSTEMS -

e CONTAINMENT u DESIGN PRESSURE (psig) 60 3 m HEAT REMOVAL CONTAINMENT SPRAY -

u ISOLATION VALVES , ~150 46 ,

e CONFINEMENT CONFINEMENT WITH FILTERED -

E EXHAUST g 9 EMERGENCY POWER 9000 28 g REQUIRED (kWe)

A 9 .

a g e m,_,,...__.,

e-, . ... .. o..... g1 . .

I E - - - _ _ -

O O O i SAFR CONTAINMENT EXCEEDS REQUIREMENTS I I

I, PRIMARY CONTAINMENT IS PROVIDED BY A LOW-LEAKAGE (0.1%/ DAY) SYSTEM SUR- l ROUNDING THE LOW-PRESSURE PRIMARY COOLANT. SYSTEM. IT CONSISTS OF A CONTAINMENT VESSEL SURROUNDING THE REACTOR VESSEL, BOTH OF WHICH ARE j ALL-WELDED SYSTEMS WITHOUT PENETRATIONS. THE REACTOR CLOSURE WITH SEAL WELDS COMPLETES THE CONTAINMENT BOUNDARY. ALL PENETRATIONS g THROUGH THE REACTOR CLOSURE ARE EITHER WELDED IN PLACE OR HAVE REDUNDANT SEALS TO MEET LEAKAGE LIMITS. PENETRATIONS SUCH AS THE GAS g PROCESSING SYSTEM ARE NORMALLY CLOSED SYSTEMS AND DOUBLY PROTECTED -

USING ISOLATION VALVES. LEAKAGE THROUGH THESE SEALS IS ON AN ORDER OF g MAGNITUDE BELOW THE DESIGN BASIS.THIS ASSURES THE DESIGN BASIS LEAK RATE SITE DOSE FROM THE WORST DBAs AND THE SSST ARE BELOW THE 10 CFR 100 gj GUIDELINES AND THE PAGs. AN ADDITIONAL LEVEL OF LEAKAGE MITIGATION IS PROVIDED BY THE REACTOR BUILDING (100%/ DAY) WHICH BACKS UP THE REACTOR COVER PORTION OF PRIMARY CONTAINMENT.

I

?

! I i '"'[,'j

, " *"""*"*"I'"""""' 37_34.g I

O O O g SAFR CONTAINMENT CONCEPT EXCEEDS REQUIREMENTS '

l SECONDARY UPPER CONTAINMENT h -

BOUNDARY g

=

~

l

, . . . . . c .., .- .. m

.. 7 I .;

IHXs lJ PUMP 1 I

h . DRHX

-~

=s  %

i l  ;

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O O O i STEAM GENERATOR BUILDING DESIGN I MINIMlZES SODIUM FIRE RISK / CONSEQUENCE l I

! I i .

1 THE SAFR IS DESIGNED SO THAT THE STEAM GENERATOR FEEDWATER AND STEAM HEADERS ARE OUTISDE THE g

l STEAM GENERATOR BUILDING. THIS ELIMINATES THE POSSIBILITY OF FAILURES IN THE SODIUM OR FEED g

WATER / STEAM LINES RESULTING IN SODIUM-WATER .

REACTION. SODIUM LEAKS ARE VERY UNLIKELY; HOW-EVER, SHOULD THEY OCCUR, THE LEAKAGE WILL BE CON- '

DUCTED TO THE CLOSED DRAIN /SWRPS TANK VAULT AND g

SMOTHERED TO PREVENT BURNING. -

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O O O a SAFR INHERENCY YlELDS LOW CORE MELT PROBABILITY l THE "INHERENCY" OF SAFR DECAY HEAT REMOVAL AND SHUTDOWN SYSTEMS HAS l

RESULTED IN SYSTEM RELIABILITY AS SHOWN ON THE OPPOSITE CHART. THE PRO-JECTED FAILURE RATES DUE TO INTERNALLY INITIATED EVENTS ARE SO LOW IN THIS l

DESIGN THAT OVERALL SYSTEM RELIABILITY IS DOM 1NATED BY SEISMIC EVENTS.

THIS SUGGESTS THAT THE SAFR DESIGN HAS ACHIEVED NEAR-OPTIMUM RELIABIL-ITY FOR THE CRITICAL SAFETY FUNCTIONS.

l i THE NORMAL HEAT REMOVAL SYSTEM (NHR) IS EXPECTED TO BE UNAVAILABLE l ABOUT ONCE IN 5 YEARS. UNDER THESE CONDITIONS, THE DIRECT REACTOR AUXIL-

) lARY COOLING SYSTEM (DRACS) WILL MAINTAIN TEMPERATURES BELOW UPSET gl i CONDITIONS (1050 F) TO PROTECT THE PLANT INVESTMENT. THE RACS SYSTEM PROTECTS THE PUBLIC HEALTH AND SAFETY BY MAINTAINING TEMPERATURES gl BELOW EMERGENCY CONDITIONS (1250 F).

gl THE HIGH RELIABILITY OF THE AUTOMATIC PLANT TRIP SYSTEMS ASSURES PROTEC-TION OF THE PLANT INVESTMENT. THE SELF-ACTUATED SHUTDOWN SYSTEM UTIL-IZES A CURIE POINT TRIGGER TO INDEPENDENTLY TRIP THE ARTICULATED SEC-

~

y ONDARY RODS.THIS SYSTEM IS INHERENT AND IMMUNE TO SABOTAGE.THE ABOVE

! PROTECTIVE FEATURES ARE SUPPLEMENTED BY INHERENT CORE RESPONSE WHICH E WILL SAFELY SHUT DOWN THE REACTOR FOR ALL CREDIBLE INITIATING EVENTS.

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O O O II i SAFR INHERENCY YlELDS jl LOW CORE MELT PROBABILITY

!I i FAILURE FREQUENCY (PER YEAR)

, INTERNAL INCLUDING SHUTDOWN HEAT REMOVAL INITIATORS SEISMIC NHR ONLY 1.8 x 101 -

NHR AND DRACS 7.2 x 10-4 -

j NHR AND RACS 1 x 10-8 -

NHR, DRACS AND RACS 3.6 x 10-a 1.1 x 10-8*

l REACTOR SHUTDOWN .

g APTS ONLY 9.7 x 10-8 -

APTS AND SASS 7.6 x 10-e _

APTS, SASS AND CORE INHERENCY 7.6 x 10 " 5 x 10-a g

l

• BASED ON GENERIC LWR FRAGILITY DATA

! NHR - NORMAL HEAT REMOVAL SYSTEM -

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! 3 i RISK ANALYSES (PRAs) OF LWRs HAVE SHOWN ACCEPT- g I ABLY LOW RISK FOR MOST PLANTS WHEN MEASURED -

AGAINST THE NRC SAFETY GOALS.

THE FUNDAMENTAL, INHERENT SAFETY FEATURES OF SAFR MAKE POSSIBLE SELF-ACTIVATING PASSIVE SAFETY i

SYSTEMS, AND CORE DESIGN HAS RESULTED IN A STEP REDUCTION IN RISK TO THE PUBLIC.

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O O O KEY FEATURES REQUIRING DEMONSTRATION I

I THE SAFR SAFETY DEMONSTRATION PROGRAM HAS TWO PHASES. THE FIRST PHASE h PROVIDES THE SAFETY DEMONSTRATIONS PREREQUISITE TO BUILDING AND LICENS-ING THE FIRST PLANT. THIS PHASE CONSISTS OF NONNUCLEAR TESTS CARRIED OUT l AT ETEC AND ANL AS WELL AS NUCLEAR TESTS CARRIED OUT AT EBR-il AND FFTF.

SOME OF THE KEY EXPERIMENTS, E.G., DECAY HEAT REMOVAL (DRACS) AT ETEC AND ATWS TESTS AT EBR-ll, HAVE BEEN COMPLETED, AND THE REMAINING TESTS ARE l

l EITHER UNDERWAY OR PLANNED. A SUBSTANTIAL PORTION ARE BEING CARRIED h l OUT UNDER THE ANL IFR PROGRAM #

THE SECOND PHASE SUPPLEMENTS THE FIRST WITH FULL-SCALE TESTS TO PROVIDE ,

I :

THE HIGHER LEVEL OF CONFIDENCE NEEDED FOR CERTIFICATION OF A STANDARD PLANT DESIGN. THESE TESTS WILL BE PERFORMED ON THE FIRST SAFR PLANT AND ll' WILL VALIDATE ON A FULL SYSTEM BASIS THE SELF-PROTECTIVE FEATURES OF THE g PLANT. '

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l . KEY FEATURES REQUIRING DEMONSTRATION l

I CODE VALIDATION TO SUPPORT FIRST UNIT LICENSING SUPPORT FOR STANDARD PLANT CERTI-FICATION g

NON-

) g NUCLEAR FIRST UNIT W ISSUE TESTS TREAT EBR-il FFTF SAFETY TEST i

DRACS PERFORMANCE 1986 1985 COMPLETED 1997 RACS PERFORMANCE 1985/1987 1986 1997 l SELF-ACTUATED SHUTDOWN SYSTEM (SASS) 1976/1990 1987 1997 RBCB PERI ORMANCE 1983/ 1997 j g 1990 .

I

! g ANTICIPATED TRANSIENTS l WITHOUT SCRAM (ATWS)

EVENTS LOSS OF FLOW WITHOUT i SCRAM 1989 1986 1986-1988 '

1997 i

LOSS OF HEAT SINK '

i WITHOUT SCRAM 1986 1997

! TRANSIENT OVERPOWER a WITHOUT SCRAM 1989 1987 1997 I

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SAFR COMPLIES WITH ADVANCED REACTOR I

I POLICY STATMENT e AT LEAST SAME DEGREE OF PROTECTION OF PUBLIC e SAFR GOAL IS TO PROVIDE SAME DEGREE OF AND ENVIRONMENT AS CURRENT LWRs . PROTECTION AS CONTEMPORARY LWRs I

e ENHANCED MARGINS OF SAFETY AND/OR SIMPLIFIED e ENHANCED MARGINS OF SAFETY AND INHERENT 4

g INHERENT, PASSIVE OR OTHER INOVATIVE MEANS TO REACTOR SHUTDOWN AND DECAY HEAT REMOVAL g ACCOMPUSH SAFETY e COMPLY WITH SAFETY GOAL POUCY e SAFR, EXCEEDS SAFETY GOALS IN NRC POLICY e HIGHLY REUABLE LESS COMPLEX PREFERABLY e INHERENT / PASSIVE SHUTDOWN AND DECAY HEAT I INHERENT OR PASSIVE SHUTDOWN AND DECAY HEAT REMOVAL SYSTEMS REMOVAL SYSTEMS I o LONGER TIME CONSTRAINTS AND APPROPRIATE INSTRUMENTATION FOR OPERATOR ACTION

• INDEFINITE STATION BLACKOUT SHUTDOWN AND DECAY HEAT REMOVAL.10s OF HRS EVENT WITHOUT SCRAM OR DECAY HEAT REMOVAL e SIMPLIFIED SAFETY SYSTEMS, REDUCED REQUIRED o FEW/ SIMPLE SAFETY SYSTEMS. NO OPERATOR ACTION I OPERATOR ACTIONS REQUIRED.4 POWER PACK 1-E POWER ONLY 80 kW e MINIMlZE POTENTIAL FOR SEVERE ACCIDENTS THROUGH e INHERENT AND DIVERSE CORE SHUTDOWN AND DECAY

• HEAT REMOVAL FOR ALL CREDIBLE INITIATING EVENTS I

INHERENCY, REUABluTY, REDUNDANCY, DIVERSITY, INDEPENDENCE e REUABLE BOP

• RELIABLE BOP FOR PLANT INVESTMENT, REACTOR ,

INSENSITIVE TO BOP g

5

• EASILY MAINTAINED e SUFFICIENT ACCESS PROVIDED e REDUCED RADIATION EXPOSURE e <25 MAN-R/ YEAR e DEFENSEIN-DEPTH e MULTIPLE RADIATION BARRIERS PROVIDED o EXISTING TECHNOLOGY OR R&D COMMITMENT e TECHNOLOGY DEMONSTRATED OR IN PLACE R&D PROGRAM j

8 '

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1 ls O O O I

I SAFR COMPLIES WITH ADVANCED REACTOR l ACCIDENT POLICY STATEMENT

! I POLICY SAFR COMPLIANCE e COMPLIANCE WITH PROCEDURAL MEETS INTENT OF REGULATIONS REQUIREMENTS AND CRITERIA OF l l i

CURRENT COMMISSION REGULATIONS

• DEMONSTRATION OF TECHNICAL COMPLIANCE INCLUDES ALL NUREG RESOLUTION OF ALL APPLICABLE- 0933 ISSUES. DECAY HEAT REMOVAL l g SYSTEM RELIABILITY ASSURED AND j UNRESOLVED SAFETY ISSUES AND ,

MEDIUM-AND HIGH-PRIORITY INDEPENDENT OF AC AND DC POWER '

l GENERIC SAFETY ISSUES

• COMPLETE PRA AND CONSIDERATION PRA DRAFT SUBMITTED TO NRC g OCTOBER 1985. NEXT UPDATE OCTOBER OF SEVERE ACCIDENT VULNERABILITIES 1986. REDUCTION OF SEVERE ACCIDENT l PROBABILITY MAJOR SAFR GOAL PRELIMINARY STAFF REVIEW INITIATED I e COMPLETE STAFF REVIEW WITH CON-

$ CLUSION OF SAFETY ACCEPTABILITY 7

1985 l E i

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O O O SAFR PROMISES COMPETITIVE ECONOMICS I I

I THIS PLOT SHOWS THE ENCOURAGING RESULTS OF THE g

DETAILED COST ESTIMATES PREPARED BY THE SAFR TEAM IN ACCORDANCE WITH DOE GUIDELINES.THESE RESULTS g

WERE CROSS CHECKED BY INDIVIDUAL TEAM MEMBERS,

, BY AN INDEPENDENT GROUP, AND WERE RECONCILED g

, WITH OTHER PLANT COSTS THROUGH DETAILED COMPAR-ISONS OF FACTORS SUCH AS CONSTRUCTION QUANTI-TIES. OUR MAIN CONCLUSIONS ARE g

• THE FIRST SAFR PLANT MAY BE ENERGY COST-COMPETITIVE WITH COAL PLANTS e MATURE SAFR PLANTS CAN HAVE CAPITAL COSTS ..

SIMILAR TO LWRs, AND ENERGY COSTS LOWER THAN l

g LWRs -

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I I O O O I SAFR PROMISES COMPETITIVE ECONOMICS (1985 DOLLARS)

STO A/ AY CYCLE

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O O O '

SAFR PROGRAM SCHEDULE I i

i I

I l OUR PLANNING INDICATES THAT DEVELOPMENT, DESIGN, CONSTRUCTION, TESTING, AND FIRST PLANT OPERATION ON A UTILITY GRID COULD BE ACCOMPLISHED ON THE SCHEDULE SHOWN OPPOSITE. THE SAFETY TESTING WILL PROVIDE A " TOTAL SYSTEM" VERIFICATION OF THE SAFR'S SELF-PROTECTIVE FEATURES AND WILL SUP-PLEMENT EARLIER " SEPARATE EFFECTS" TESTING TO PROVIDE A BASIS FOR THE CERTIFICATION OF THE SAFR STANDARD DESIGN. THIS WILL GREATLY FACILITATE l AND EXPEDITE THE LICENSING OF SUBSEQUENT SAFR UNITS BY FOCUSING THE LICENSING REVIEW AND HEARINGS SOLELY ON SITE-SPECIFIC ISSUES.

THE SAFR FIRST PLANT APPROACH ALSO WILL DEMONSTRATE OPERATION AND MAINTENANCE IN A UTILITY ENVIRONMENT AND PROVIDE A BASIS FOR CONFIDENCE -

l IN THE PREDICTABILITY OF PROJECT COSTS AND SCHEDULES. REASONABLE SUC-CESS ON THE ABOVE PROGRAM WILL BE INVALUABLE IN ESTABLISHING UTILITY /

l USER / INVESTOR CONFIDENCE WHICH IS PREREQUISITE TO A PENETRATION BY SAFR l j OF THE DOMESTIC ELECTRICAL GENERATION MARKET. -

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I SAFR PROGRAM SCHEDULE N N ST N N N el N N N N N ST . N N 1 R W I  : - t=A g

r- 7: ''

CONSTRUCTION EST T OPE T

$lTE v I  !

ER l RESULTS

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I COutt0N ACTMTES CONCEPTUAL DE54GN PREllM6 MARY / DETAIL DESIGN / DESIGN I

g ll l I

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! g SAFR 8 ICATION E ER l

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' I T L S ARO E AT N CONSTRUCTION /i I

T PL NT R 8D ' f5AR FOR $URSEQUENT PSAR PLANTS h$

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!I SAFR FIRST PLANT FUNDING APPROACH lI i

l 8 STATUS - UNDERGOING FINAL CORPORATE REVIEW ll j

e MAIN FEATURES - BY CAPITALIZING ON ECONOMIC Y^.LUE OF PLANT, PLAN PROVIDES l e COST TO GOVERNMENT BEFORE OPERATION < $300M l e ULTIMATE NET COST TO GOVERNMENT ~ 0

! g

• MANAGEABLE ANNUAL FUNDING LEVELS l 5 e REASONABLE RETURN TO OPERATING COMPANY ,

l l

• PARTICIPANT RISKS COMMENSURATE WITH CONTROL AND BENEFITS I s -

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LI -- - - - - -- - -- --

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SUMMARY

l l

l 4 I

• SELF-PROTECTIVE FEATURES ASSURING RISKS R LWRs OF SAME l

i l TIME FRAME g

• ECONOMICALLY COMPETITIVE WITH COAL AND LWRs

• PLANT SIZE, CONSTRUCTION SCHEDULE COMPATIBLE WITH

! UTILITY PLANNING NEEDS e j

l

• USER FRIENDLY l

• UTILITIES INTERESTED IN PROGRAM, UNREADY TO ASSUME l

j MAJOR FINANCIAL ROLE

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! Sodium ADVANCED FAST REACTOR ll AGENDA l

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!E e INTRODUCTION e DESIGN OVERVIEW 1

e SAFETY GOALS / FEATURES  :

e ECONOMICS I e FIRST PLANT PLAN i

l  !

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'4-7a-a I . - _ _ _ _

n _ O O i

I SAFR CONCEPT AIMS AT ALLEVIATION OF g MANY NUCLEAR PLANT CONCERNS lI o BY PROVIDING B . " NATURAL" SAFETY

!i . _ _ mC _ Ce ,

e SHORTER, PREDICTABLE CONSTRUCTION SCHEDULES .

• IMPROVED AVAILABILITY u

I $

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O O O a,

! a i FUNDAMENTAL PROPERTIES OF LIQUID METAL REACTORS TRANSLATE l INTO DESIGN ADVANTAGES THE SUPERIOR FUNDAMENTAL PROPERTIES OF LIQUID METAL REACTORS

! TRANSLATE INTO VERY FAVORABLE LMR PLANT CHARACTERISTICS. BY CAPI-TAllZING ON THESE PHYSICAL PROPERTIES, LMRs CAN BE DESIGNED SUCH j i THAT N ATUR AL FORCES (E.G., GRAVITY, THERMAL EXPANSION) ESTABLISH S AFE i

AND STABLE CONDITIONS FOLLOWING ANY CREDIBLE REACTOR UPSET OR g ACCIDENT. ADDITIONALLY, THE PROPERTIES OF SODIUM ALLOW PLANT EFF1- E

! CIENCIES TO APPRAOCH THOSE OF FOSSIL PLANTS AND PERMIT USE OF CON-VENTIONAL HIGH SPEED STEAM TURBINES. j 1

3 R

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O O i FUNDAMENTAL PROPERTIES OF LIQUID METAL l I~ REACTORS TRANSLATE INTO DESIGN ADVANTAGES l

! PLANT CHARACTERISTICS

!I COOLANT l HIGH THERMAL EFFICIENT, LOW PRESSURE COOLANT SYSTEM I

il CONDUCTIVITY HIGH PLANT EFFICIENCY l

l 3 LOW VAPOR PRESSURE 8 CONVENTIONAL FOSSIL l  :

! NONCORROSlVE OF STEAM TURBINES

!$ STRUCTURAL MATERIALS N PASSIVE DECAY HEAT j HIGH BOILING POINT REMOVAL STRONG NATURAL LONG " GRACE PERIODS" g

l CONVECTION l 9 LOW RADI ATION EXPOSURES I 1, j f 2 i

i @ =:: n r l __ _

i l O O O i

. I LIQUID METAL REACTORS AROUND THE WORLD l I 3

i THIS MONTAGE PICTURES SEVERAL OF THE WORLD'S SODIUM-COOLED REAC-TORS. THE SODIUM REACTOR EXPERIMENT (SRE) (DESIGNED / CONSTRUCTED / l OPERATED BY ROCKWELL INTERNATIONAL) WAS THE FIRST REACTOR IN THE

! WORLD TO SUPPLY ELE ^TRICITY TO A UTILITY GRID.THE 20-MWe EXPERI- l MENTAL BREEDER REACTOR-il HAS OPERATED AT HIGH CAPACITY FACTORS FOR OVER 20 YEARS, AND THE FAST FLUX TEST FACILITY HAS OPERATED WELL SINCE ITS STARTUP IN 1982. THE 250-MWe PHENIX, THE 1200-MWe SUPER- t l

PHENIX, AND THE 250-MWe PROTOTYPE FAST REACTOR (PFR) ARE OPERATING l

) SUCCESSFULLY ON FRENCH AND UNITED KINGDOM UTILITY SYSTEMS. THE RUSSIAN BN 350 HAS BEEN GENERATING ELECTRICITY AND PRODUCING j j DESALINIZED WATER FOR OVER A DECADE. ANOTHER RUSSIAN PLANT, THE

] 600-MWe BN 600, WENT CRITICAL IN JANUARY 1982 AND WAS TIED INTO THE l

GRID TWO MONTHS LATER. ,

l b

l S ete=- dit) -->-- 14-75-1 3

l USA JAPAN I SRE JOYO 5-*

! LIQUlO METAL REACTORS I l - '. ' l '

l. AROUND THE WORLD , -

PWR ANTS OPERATED

." *s!OSU"TkiggLE i

' e e A R GH N i HNPF 1! ..

• TECHNOLOGY SUPPORTED BY >$78 OF U.S.

DEVELOPMENT AND DEMONSTRATION MONJU j , , .

w UNITED KINGDOM .,

PFR _

- USSR FRANCE l FFTF PHENIX BN-350 g

ll WEST GERMANY BN j SNR-sa . -600 s . .

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OBJECTIVES OF DOE SPONSORED SAFR PROGRAM I

4 .

E I

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• ASSESS ELECTRIC INDUSTRY NEEDS

• DEVELOP SAFR TO MEET THESE NEEDS

• ESTABLISH BASIS / APPROACH FOR STANDARD l -

DESIGN CERTIFICATION LEADING TOWARDS COMMERCIAL INTRODUCTION h

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O O O l

SAFR PROGRAM TEAM

THE SODIUM ADVANCED FAST REACTOR (SAFR) TEAM MEMBERS AND THEIR i RESPONSIBILITIES ARE SHOWN IN THE UPPER PORTION OF THIS CHART. THIS i OUTSTANDING GROUP HAS COLLABORATED FOR MANY YEARS ON LMR WORK, l

WITH EXCELLENT RESULTS. DOE HAS ORIENTED THE LMR BASE PROGRAM TO PROVIDE STRONG SUPPORT TO THE INNOVATIVE DESIGN EFFORTS. WE HAVE l DEVELOPED A DETAILED, COORDINATED SET OF WORK PLANS WITH OTHER BASE PROGRAM ORGANIZATIONS, AND THE RESULTS ARE SHOWN IN THE BOT-l TOM PORTION OF THIS CHART. THIS SECTION SHOWS THE NUMEROUS TASK-l AREAS IN WHICH BASE PROGRAM WORK IS BEING DONE BY LABORATORIES AND i OTHER CONTRACTORS FOR SPECIFIC INCORPORATION IN THE SAFR DESIGN PACKAGE.

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Hockweilinternational seassesviess%assessanaues noca.idyn. o wiS

• 15-34-0 _

O O O usesirsg Tj P.its a l

NdliD W _

Presentation To Nuclear Regulatory Commission October 9,1986 i

G GENERAL ELECTRIC i NUCLEAR SYSTEMS TECHNOLOGY OPERAT10N 16 443411

i o o o I

mm i n as =

PM D I W S ..

i

'i PRISM DESIGN AND PROGRAM i

. . . Dr. J.S. Armijo Manager, General Electric Nuclear Systems and Technology Opeiation ,

l i

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scaue2 l

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O O O

\ PRISM AND THE NEEDS OF THE MARKET l

C MME" '^ '

l Innovative Program CI C 20

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l Reactor Certification i

l 1

)

\

p INITIATE SAFETY TEST (DOE NRC,1NDUSTRY, UTILITY) l h e CONCEPTUAL DESIGN,36 MONTHS i

@ e OBJECTIVES: .

- CHAR ACTERISTICS RESPONDING

@ / TO MARKET NEEDS

.i g - COST COMPETITIVE

- ENH ANCED SAFETY

@ -IMPROVED LiCENSABILITY/ CERTIFICATION gg -HIGH AVAILABILITY

- COMMERCIAL ACCEPTABILITY DOE INITI ATES COMPETITIVE PROGR AM GE INITI ATES INNOVATIVE DESIGN PROGRAM i e OBSTACLES TO NUCLEAR EVIDENT

• SOLUTION - NEW APPROACH -SAFETY TEST 1

86 018 01

O O O

} PRISM DESIGN APPROACH

~

i ,

J -

e COMPACT POOL-TYPE REACTOR MODULES SIZED TO ENABLE:

- FACTORY FABRICATION WITH MINIMUM SITE INSTALLATION LABOR

- ECONOMICAL SHIPMENT TO INLAND AS WELL AS WATER-SIDE SITES

- ECONOMICAL FULL-SCALE TEST FOR CERTIFYING STANDARD DESIGN

-l e CAPABILITY FOR INCREMENTAL POWER BLOCK ADDITIONS AND MODULE REPLACEMENT e SAFETY-RELATED EQUIPMENT LIMITED TO REACTOR MODULE AND SERVICE SYSTEMS 8644343

1 l

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REACTOR MODULE REACTOR Q CONTROL DRIVE ASSEMBLIES CLOSURE REACTOR SUPPORT

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PRISM NUCLEAR STEAM SUPPLY SYSTEM 1 . . , , , , .

i NUCLEAR NON-SAFETY SAFETY 4 k RELATED RELATED i

! RVACS j MM STACKS t

! I I I I III I II

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REACTOR MODULE kh, E 2 g

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86-443 4)

O Q O .

DESIGN BASIS REQUIREMENTS

• COMPLY WITH TOP LEVEL REGULATORY REQUIREMENTS ON RADIATION EXPOSURE (10CFR20,10CFR50, 10CFR50 APP.1,10CFR100, 40CFR190) -

• PROVIDE THE FOLLOWING HIGHLY RELIABLE SAFETY-RELATED SYSTEMS ON EACH REACTOR MODULE

- SHUTDOWN SYSTEM

-SHUTDOWN HEAT REMOVAL SYSTEM

- CONTAINMENT 16 443-21

! O O O INHERENCY-SAFETY OBJECTIVE e REACTOR SHUTDOWN COOLING AND CONTAINMENT OF CORE FISSION PRODUCTS CAN BE PERFORMED PASSIVELY, DEPENDENT ONLY ON REACTOR MODULE GEOMETRY AND MATERIAL SELECTION, FOR THE FOLLOWING EVENTS:

- LOSS OF PRIMARY, INTERMEDIATE A D STEAM / WATER COOLANT FLOW WITHOUT SCRAM

- LOSS OF NORMAL HEAT SINK OR INTERMEDIATE SODIUM COOLANT WITHOUT SCRAM e REACTOR AND REACTOR C00LANT SYSTEMS CAN SAFELY ACCOMMODATE A SINGLE R0D WITHDRAWAL WITHOUT SCRAM i

l RESPONSIVE TO ADVANCED REACTOR POLICY STATEMENT i

8644344

O O O RESPONSIVE TO ADVANCED REACTOR POLICY STATEMENT l 1. HIGHLY RELIABLE AND LESS COMPLEX

• PASSIVE SIMPLE AIR COOLED SYSTEM

! SHUTDOWN AND DECAY HEAT REMOVAL USING NATURAL CIRCULATION SYSTEM

• SHUTDOWN SYSTEMS ARE REDUCED IN

! NUMBER AND SIMPLIFIED WITH INTERNAL

! DIVERSITY 4

2. COMPLY WITH SAFETY GOAL POLICY

• PRA ASSESSMENTS USED TO GUIDE DESIGN STATEMENT AND SHOW COMPLIANCE TO SAFETY GOALS I

3. ENHANCED MARGINS 8

• LOW TEMPERATURE (875 F CORE OUTLET) ,

1000 PSIG SATURATED STEAM CYCLE

• LOW CORE POWER DENSITY 13 KW/ft METAL FUEL

• SEISMIC ISOLATION OF REACTOR MODULE i

I 1

]

mms

l O O O i

RESPONSIVE TO ADVANCED REACTOR POLICY STATEMENT i

4. USE SIMPLIFIED, PASSIVE, INHERENT e RVACS IS INNOVATIVE PASSIVE SHUTDOWN j OR OTHER INNOVATIVE MEANS TO HEAT REMOVAL SYSTEM i ACCOMPLISH SAFETY

!

• SEALED INERT CONTAINMENT WITH NO ACTIVE PENETRATIONS DURING OPERATION

• METAL FUEL CORE INCORPORATES INHERENT SAFETY MARGINS AND NEGATIVE REACTIVITY FEEDBACK

5. MINIMlZE THE POTENTIAL FOR e HIGHLY RELIABLE SHUTDOWN SYSTEMS AND SEVERE ACCIDENTS SHUTDOWN HEAT REMOVAL SYSTEMS

• PROTOTYPICAL MODULE SAFETY TEST TO DEMONSTRATE CAPABILITY TO ACCOMMODATE SCRAM FAILURE o HIGH POTENTIAL FOR IN-VESSEL COOLING OF DAMAGED METAL FUELED CORE eS4um I

O O O PRISM SHUTDOWN HEAT REMOVAL SYSTfMS I

i j

TURBINE STEAM FLOW CONTROL s

50010M r 5M

i E . -.

ggh '

j AIR

,, - m g

[ ,v i {

{~

f s

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e

. ;,4

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! 3 r

+ r () d I , ,

Ti CONGENSER f i HEAT n! (  ; II

?

M _

, . DEAERATOR 6T6- ANDfW HEATERS r-air ? .

REACTOR VESSEL AUXILIARY COOLING AUXILIARY COOLING NORMAL CONDENSER SYSTEM (RVACSI SYSTEM (ACS) COOLING

-SAFETY GRADE -NON SAFETY GRADE -NON SAFETY GRADE ilCA 4347 I

._ _. _ _ _ _ _ _ _ _______J

O O O ,

r l PRISM SHUTDOWN HEAT REMOVAL LOSS OF IHTS. COOLING WITH SCRAM FROM FULL POWER 1

1300 l l l l l l l l ji DESIGN LIMIT (level D) l 1200 - -

RVACS ONLY

<0NE PER LIFETIME i 1100 -

g -

! 7-O# ^ "

1000 - -

TEMPERATURE (8) F 900  %

N AUXILIARY COOLING (ACS)

% < TEN PER LIFETIME 800 -  % -

700 -

%~~_ -

600 I I I I I I I I I

O 10 20 30 40 50 60 70 80 90 100 TIME (Hours) 1644308

CAPITAL COST COMPARISON (19865)

COST

($/KWe) 10,000 &CRBR g ,  ;  ; i 9000 -

$ MONJU 8000 -

7000 - -

6000 - -

5000 INITIAL -

. PRISM l MODULE t 4000 - -

l l

l SUPER 3M -

LEAD PHENIX' OPOWER PRISM BLOCK LWR l 4 2000 -

f*'*===,. o " , , , ,

' ' "' , U, A a' w n Coal l.g n,.

' '. w . , .....: .

e .,. ". -

. . .o , ?

g . ,.,: -.t; .: y.:;. c:

p.e,..: .g, . ,. .. '.sp,. M.

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~ -

s ;,,,; g.' .. . 9..A* '='* hgw. e y 1

! 1000 I I I I I 200 400 600 800 1000 1200 1400 PLANT CAPACITY (MWe)

I 86 'f.30 02

BUSBAR COST COMPARISON (19865) 90 1

I l 80 - -

70 - -

LEAD ,

60 -

$ POWER Coal PRISM _

COST BLOCK LWR (Mills /kWh) =

50 k,,,,, , ' % ,,

g.c. ...._.  ; ::.

=.==,.,,,,-

c

,~

,o m ,<,,, ,,, ,,, ,,, ,,, ,,,, ,,,,, ,,,

m l l;. . .y . .;p :. . .

4g l

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g.@%.t.?

= w PRISM COMMERCIAL PLANTS ~

. .;g 4 ,;,;, g ~..c ~~ ~ u.

-l

,, ,,,,, L '"".""""" . . . . ..- . . W - - O J 30 - -

I I I I I I I I l I 20 400 600 800 1000 1200 1400 PLANT CAPACITY (MWe)

PRISM WILL BE COMMERCIAL AFTER SUCCESSFUL SAFETY TEST 86-443-11

! O O O PRISM COMMERCIALIZATION PLAN i

1 I

l DEMONSTRATE PERFORMANCE !)

! 0F KEY R&D i COMPONENTS a

DEMONSTRATE FIRST  ?

PROT YPE PRISM SAFETY TEST  : COM RCIAL j PERFORMANCE DEMONSTRATE POWER OPERATION l POWER d OPERATION FUEL CYCLE DEMD {

, n

! l$0NSTRATE METAL FUEL FAB l CYCLE FUEL REPROCESSING & WASTE MANAGEMENT h l 1986 1989 1992- 1995 1998 l

BiM43-12

O O O P R S M ..

i PRISM APPROACH TO LICENSING -

j l

. . . Dr. G.G. Sherwood l Manager, General Electric Nuclear i

\ Safety & Licensing Services 1

l I

-l i

86443-13

O O O APPROACH TO SAFETY AND LICENSING -

• lNHERENT SAFETY OF PRISM

- PASSIVE SHUTDOWN l - PASSIVE DECAY HEAT REMOVAL ,

l l

• SAFETY TEST

- INHERENT FEATU RES PERMIT FU LL-SCALE PROTOTYPIC TEST I

-MARGINS ALLOW OPERATION

• CERTIFICATION

- SAFETY TEST WILL SUPPORT CERTIFICATION OF STD. DESIGN ,

- CERTIFICATION WILL GIVE USER AND PUBLIC HIGH CONFIDENCE i

86 443-15

O O O PRISM SAFETY TEST ACTIVITIES

e DEMONSTRATE OPERATING AND INHERENT SAFETY CHARACTERISTICS

- BASIS FOR DESIGN CERTIFICATION BY NRC I e TESTSWITHIN AND BEYOND DESIGN BASIS EVENTS

- CONVENTIONAL TESTS: -

PRE-OPERATION, BASELINE ISl, HOT FUNCTIONAL, FUEL LOAD, STARTUP

- BENCHMARK TESTS TO VERIFY INHERENT RESPONSE:

! REACTIVITY FEEDBACKS, HEAT REMOVAL i - SAFETY TESTS I

WITHIN DESIGN BASIS -WITH SCRAM ROD WITHDRAWAL, LOSS OF FLOW, LOSS OF HEAT SINK ,

BEYOND DESIGN BASIS-WITHOUT SCRAM ROD WITHDRAWAL, LOSS OF FLOW, LOSS OF HEAT SINK, LOSS OF ELECTRICAL POWER, DEGRADED AUXlLIARY COOLING (RVACS-WITH SCRAM) e TESTWILL ALSO PROVIDE:

- DEMONSTRATE REACTOR MODULE FACTORY FABRICATION AND ASSEMBLY

- DEMONSTRATE MODULAR CONSTRUCTION 86-4434 7

O Q

8ee 5

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0

O O O PRISM TEST AIR DUMP HEAT EXCHANGER SYSTEM HEAT

• STANDARD REACTOR

• 4 REJECTION m

sonium b '""'

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_ _ AIR DUMP d{ g, {f HEAT EXCHANGERS REACTOR MODULE 86-342 04

i l CONTENTS OF DRAFT PROPOSED APPENDIX S i

• SIMILAR TO APPENDIX 0 i

olNTENDED FOR FINAL DESIGN ONLY j

• PROVIDES FOR PROTOTYPICAL SAFETY TEST

• CONSISTENT WITH:

i

4

- NRC SEVERE ACCIDENT POLICY STATEMENT

- NRC BACKFIT RULE -

l - PENDING LICENSING REFORM i

LEGISLATION

- NRC POLICY STATEMENT ON

! STANDARDIZATION'(Under Consideration) AND INDUSTRY COMMENTS THERETO i

- ADVANCED REACTOR POLICY STATEMENT 86-44 M 8

O O O

~

ADVANTAGES OF PROPOSED APPENDIX S l

l

) eREDUCES UNCERTAINTIES ASSOCIATED WITH APPLYING , ,

j APPENDIX 0 TO ADVANCED REACTORS

• ENCOURAGES DEVELOPMENT OF REACTORS WITH. FULLY DEMONSTRATED SAFETY CHARACTERISTICS l

l e SPECIFICALLY IDENTIFIES ROLE OF SAFETY TEST i

o ESTABLISHES AN EFFICIENT ALTERNATIVE FOR CERTIFYING ADVANCED REACTORS l

l 1

8tM4319

O O O PRISM CERTIFICATION SCHEDULE ,

FISCAL YEARS 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 STANDARDIZATION CONTAINMENT COMMISSION REPORTS ACCIDENTS SUBMIT PSER PSID v

CONCEPTUAL DESIGN REVIEW SAFETY SAFETY SAFETY TEST TEST TEST DESIGN PLAN COMPLETE V V V SAFETY TEST APPROVAL PROPOSE RULE STD DESIGN SAFETY hf RULE APPROVED REPORT SER TEST DESIGN v v v v REPORT v CERTIF1-CATION STANDARD MODULE CERTIFICATION 86443-22

O O O SAFETY AND LICENSING

SUMMARY

1

• SOUND LICENSING PROGRAM FOR CERTIFICATION

!

• NRR INTER ACTIONS (85-86) -

1 1

- AGREED ON APPROACH

-TECHNICAL MEETINGS UNDERWAY

• PROPOSED APPENDIX S SUBMITTED i

• NRR MANPOWER MAY STILL BE ISSUE '

)

!

• PRISM CONCEPT PROVIDES UNIQUE OPPORTUNITY i

l i

l!

l

< m l MODULAR HTGR l

DESIGN OVERVIEW l PRESENTED TO THE NRC COMMISSIONERS OCTOBER 9,1986

~

T. E. NORTHUP, GENERAL MANAGER POWER REACTOR PROGRAMS GA TECHNOLOGIES INC.

[PUFFENBUR]Il3 1 2-0CT-8 8

O O O Tile MODULAR llTGH c.m,~, -

i

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GRAPillTE MODEllATOR/CollE STilUCTUllE

CEllAMIC CLAD FUEL

~

lilGil TEMPEllATURE/IllGil EFFICIENCY .

LOW IIADl0 LOGICAL WASTES, LOW W0llKER EXPOSUIlE l e SPECIAL MODULAll FEATUllES INilEllENT, PASSIVE SAFETY SNIALL UNIT llATING

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O E F A S Y N E - T S di V - E M T I S F E A A T U S S S C A Y A P C S V E I L Y T R B h O U F O P A S N I S D R - E E E l l R T L U E E S W O l i S S P A S I C Y L l N N B OT i - G O I A U T F P l l C O F A O l l - R N A O O D L T I E T E U A A N R R D E E E O P P i l O O M T T T T E l U U U i 0 l 0 0 T i T l i l i I T I T I W W W 2 g

                                                          '2li.

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l O O O r ,

        . . ~ , .

1 MHTGR SAFETY AND LICENSING APPROACH RESPONSE' { TO ADVANCED REACTOR POLICY j POLICY OBJECTIVE MiTGR o ENHANCED MARGINS OF SAFETY o MEETS PAG LIMIT AT EAB. l (NO OFFSITE SHELTERING / EVACUATION)'USING PASSIVE SAFETY FEATURES. o EARLY NRC INTERACTIONS o TO DATE: 12 NRR MEETINGS & ' 2 ACRS MEETINGS 8 SUBMITTALS INCLUDING o LICENSING PLAN - o TOP LEVEL CRITERIA o 4 VOL. PSID o FUTURE INCL PRA AND TECl-NOLOGY PLAN LEADING TO PSER AND LICENSABILITY

                        ,                            STATEMENT                                           '
 ) (                                                                                                 )

[PUFFENBUR]I26 1 2-0CT-86

O O O

   <                                                                                                                   m mas ou~,... sem MHTGR DESIGN RESPONSIVE TO ADVANCED REACTOR POLICY POLICY OBJECTIVE                     MiTGR DESIGN i

o LESS COMPLEX SHUTDOM4 o SHUTDOM4 AND DECAY HEAT REMOVAL - LARGE NEG. TEMP COEFF 1%O DIVERSE, GRAVITY CONTROL ROD AND 4 SHUTDOAN SYSTEMS l o DECAY HEAT REMOVAL 1%O, INDEPENDENT ~

,                                                                            FORCED CORE COOLING                       -

SYSTEMS SEPARATE, PASSIVE ' NATURAL CIRCULATION

                                                                           ' HEAT REMOVAL j                         o LONGER TIME CONSTANTS -       o    HIGH HEAT CAPACITY AND LON I

FOR ACCIDENT MANAGEMENT POAER DENSITY PROVIDE DAYS DEFORE FUEL TEhPERATURES PEAK WELL BELON LIMITS ( ) ['tJFFENBUR]t26 2 2-0CT-86

O O O r , mm= . .. m m . MHTGR DESIGN RESPONSIVE TO ADVANCED REACTOR POUCY (CONTD) POLICY OBJECTIVE MiTGR DESIGN o SIWLiflED SAFETY SYSTEMS o NO OPERATOR ACTION REQUIRED TO REDUCE OPERATOR ACTIONS FOR SIUIDOAH A m HEAT REWVAL o MINIMUM POTENTIAL FOR o NUCLEAR REACTION SELF SEVERE ACCIDENTS TERMINATES AS 00RE HEATS UP o LOSS OF COOLANT ACCEPTABLE 1 o COOLANT CHEMICALLY AND NEUTRONICALLY INERT o NO PHASE CHANGES ~ o RELIABLE BOP OR SAFETY o NO ACTIVE POWERED SAFETY SYSTEMS INDEPEt0ENT SYSTEMS , o INSENSITIVE TO BOP TRANSIENTS o LON MAINTENANCE A?O o HIGH QUALITY FUEL RETAINS PERS0tNEL EXPOSURES FISSION PRODUCTS o LON HELILN AND PRIMARY CIRCUIT ACTIVITY - EASY ACCESSIBILITY

                                                                                                                                                     )

(PUFFENBUR]I26 3 2-0CT-86

O O O r m j me c. . sua MHTGR DESIGN RESPONSIVE TO ADVANCED REACTOR POLICY (CONTD) POLICY OBJECTIVE MITCR DESIGN o DEFENSE-IN-DEPTH WITH o BARRIERS MULTIPLE BARRIERS AND - TRISO COATED PARTICLES REDUCED ACCIDENT - FUEL Eh0EDDED WITillN POIENTIAL AiO GRAPillTE CONSEQUENCES - STEEL PRESSURE VESSEL

                                                   - BELON GRADE STRLOTURES o   PREVENil0N#1TIGAT10N filGILY RELIABLE, OPERATOR FRIENDLY,
                                                   - FORGIVING DESIGN

- NEGLIGlBLE RELEASES l -

o PROVEN TECINOLOGY o USES GAS COOLED EXPERIENCE MAGNOX, AGR, AVR, THIR, j PEACH BOTT W 1, FSV , o FUEL PARTICLE PERFORMANCE DEMONSTRATED o LON ENRICHED TilROVAWAY FUEL AVOIDSPROLIFERATION/ ! REPROCESSING ISSUES l L ) [PUf7EHOUR]I26 4 2-0CT-86

              .o-      e                                         - - - .

O N BRIEFING ON ADVANCED REACTORS . FOR THE NUCLEAR REGULATORY COMMISSION l l GAS-COOLED REACTOR ASSOCIATES OCTOBER 9,1986 l

O O O NUCLEAR PLANT SIZE PREFERENCE , 19 18 17 - 1 6 - ----------- @ ! 15- ----------- i 14 i-13 cn 12 -------------- 11 tr 10 8 - -

      $          7 3          6                                         E                   -

z 5 ---------- 4 3 T I i 1- - O , I < 200 200-400 400-700 700-1000 > 1000 I PLANT SIZE PREFERENCE (MWe) 1 i

i

SUMMARY

OF UTILITY / USER REQUIREMENTS FOR MODULAR HTGR DESIGN CPITERIA UTILITY / USER REQUIREMENT EQUIVALENT ANNUAL AVAILABILITY TOTAL OUTAGE 20% MAXIMUM OVER LIFETIME SCHEDULED OUTAGE 10% MAXIMUM OVER LIFETIME PI. ANT INVESTMENT PROTECTION UNSCHEDULED OUTAGE 10% MAXIMUM OVER LIFETIME OUTAGES > 6 MONTHS 10% MAXIMUM OF UNSCHEDULED OUTAGES EXPECTED VALUE OF LOSS < ANNUAL INSURANCE PREMIUM PROBABILITY OF EXCEEDING SAFETY RELATED DESIGN LIMIT < 10-5/ PLANT-YR SAFETY AND LICENSING CRITERIA OVERALL CRITERIA EXISTING NRC/ EPA DOSE AND RISK CRITERIA EMERGENCY PLANNING CRITERIA NO SHELTERING OR EVACUATION REQUIRED SITING PARAMETERS EXCLUSION AREA BOUNDARY RADIUS 425 METERS SEISMIC (GROUND ACCELERATION) .3 g SSE/.15 g CBE FUEL CYCLE ENRICHMENT LEVEL LOW, < 20% SPENT FUEL MANAGEMENT ONCE-THROUGH THROWAWAY I ECONOMIC GOALS BUSBAR POWER COST AT LEAST 10% ADVANTAGE OVER I COMPARABLY SIZED ADVANCED COAL PLANTS INSTALLED CAPITAL COST < 2000$/KW (1986 DOLLARS) i

1

      ~'N (O                                                                                  :

l MHTGR DEMONSTRATION PROJECT OBJECTIVES e DEM'ONSTRATE THE LICENSING PROCESS USING THE CRITERIA AND METHODOLOGY ESTABLISHED FOR THE MHTGR PROGRAM AND SUPPORT DESIGN CERTIFICATION EFFORT, AS REQUIRED, FOR THE STANDARD MHTGR e DEMONSTRATE PLANT PERFORMANCE CHARACTERISTICS O e DEMONSTRATE CRITICAL MAINTENANCE ACTIVITIES AND SYSTEM / COMPONENT RELIABILITY l -, o ESTABLISH BASIS FOR COMMERCIAL PLANT COST AND SCHEDULE PLUS FOSTER THE DEVELOPMENT OF A VENDOR / SUPPLIER INFRASTRUCTURE 4 l e ESTABLISH UTILITY / USER / INVESTOR CONFIDENCE TO BUY COMMERCIAL PLANTS

O

1 I l PROJECT DEFINITION STUDY OBJECTIVES e SCOPE & LAYOUT e LICENSING APPROACH e SITING OPTIONS e TESTING PROGRAM e TECHNOLOGY REQUIREMENTS e PROGRAM COST ! e PROGRAM SCHEDULE l l I O

O I MGR TARGET PROJECT COSTS i BREAKDOWN THROUGH 1997 (TOTAL - $800 M) i l TECHNOLOGY DESIGN & IC. (8*0%) (10.6%) CONTINGENCY (13.3%) DEMO. PLANT DESIGN & LIC. FUEL AND O&M ' Z (7.0%) , OWNER'S COST (6.8%) FIELD OFFICE (4.1 %) (28.9%) PLANT HARDWARE O

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MGR DEMONSTRATION PROJECT PHASES PHASE I - PROJECT DEFINITION (THRU 1988) PHASE IIA - PRELIMINARY DESIGN AND LICENSING (1989- 1990)

 ;                                                         PHASE IIB -

DETAILED DESIGN AND LICENSING /LONG-LEAD MANUFACTURING (1991 - 1992) PHASE III - MANUFACTURING, CONSTRUCTION AND TESTING (1993-1997) PHASE IV - COMMERCIALIZATION 4 O 1

l l

l MGR DEMONSTRATION PROJECT PHASES ] PMASE I - PROJECT DEFINITION (THRU 19881 e CONCEPTUAL DESIGN AND EVALUATED ECONOMIC BAS!S ! ESTABLISHED FOR REFERENCE PIANT e LICENSING BASIS ESTABLISHED THROUGH PSID REVIEW AND NRC ISSUANCE OF LICENSABILITY STATEMENT

 ;                                                e   CONCEPTUAL                         DESIGN AND                                             BASELINE COST ESTIMATE ESTABLISHED FOR DEMONSTRATION PLANT e   PROJECT DEFINITION, INCLUDING COSTS, LICENSING P!.AN ,

1 TEST PLAN & SITE ESTABLISHED e PROJECT STRATEGY PIAN , INCLUDING COST / RISK SHARING AND

 ,                                                    MANAGEMENT ARRANGEMENTS ESTABLISHED e   UTILITY MOST                                 (UPP & UPC)                                   AND SUPPORT ARRANGEMENTS ESTABLISHED i                                                  e   VENDOR / SUPPLIER /AE                                                              TEAMING        (PSC)    AND SUPPORT ARRANGEMENTS ESTABLISHED

e DETAILED PROJECT PLAN (SCOPE, COSTS, SCHEDULE)

ESTABLISHED AS BASIS FOR PROJECT COMMITMENT j PMAce 7'A - PRettv' NARY DESIGN AND LICENSING (1989 -19901 i . l e PRELIMINARY SAFETY ANALYSIS REPORTS SUBMITTED FOR ! REFERENCE PIANT AND DEMONSTRATION PIANT e FIXED PRICES ESTABLISHED FOR MAJOR COMPONENTS AN0/OR ' SYSTEMS IN DEMONSTPATION PLANT - COMMIT LONG-LEA 0 ' MANUFACTURING PHASE !!B - DETA* LED DESIGN AND 1: CENSING /LONG-LIA0 MANUTACTTRING (1991 - 19921 o PRELIMINARY DESIGN APPROVAL ISSUANCE FROM NRC FOR REFERENCE PLANT e CONSTRUCTION PERMIT ISSUANCE FROM THE NRC FOR DEMONSTRATION PLANT i PHASE !!! - MANUFACTURING. CONSTRUCTION AND TESTING (1993 -199

e MANUFACTURING AND CONSTRUCTION COMPLETF" i e OPERATING LICENSE ISSUANCE FROM THE NRC FCR DEMONSTRATION PIANT e STARTUP AND DEMONSTRATICN TESTS COMPLETED PMASE I'? - COMMERCIALIZATION 4

j e COMMERCIAL PLANT FINAL DESIGN ESTABLISHED THRCUGH FSSAR ! REVIEW BY NRC AND ISSUANCE OF FDA AND DESIGN i CERTIFICATION e COMMERCIAL ORDER (S) COMMITTED 1

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pg i y \ l 1984 1985 1988 1989;1990 g 1991 g 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 1986l1987 REFE8ENCE PLANT _DLYEL, , C000 CEPT CoseCEPTUAL PRELBASIAny l PL ANT DESIGN EV AL. DESaGet DESeQge

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\ ICE S NG PAEAPPL8 CAT 8086 PitEL98HedARY REVDEW S A DN t , etent vDP I l TECHNQLOGY_ DEVELOPMENT TEC80eOLOGY .agte naars puntacoesp. j gt j DEVEL.PLA808 ,,tc castensa anaseants Coeset. N!'"' P ce pN , i DEMONS _lBAIK)N PLANI I I e 4 a l PROJECT DEF*N/ OEF NicoseCEPT PRELIMINARY DES 4006 FesAL e DESIGN /LICEN. DESsess LICENS8860 M S4GB8/L8C. , l ,';;,d:' dU=a

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                                                                                                                                                                  /CoseSTnUCTIOes LONG-TERM DEaso TESTS                         OPER A TION 4                                                                                                                                                                                         ,,,,,,,
                                                                                                                                                                                                                                 .e e.

COMMERCIAL PLANI "l'" "I' ""** ' STANDARD NI DESIGN /LICEN. F100AL DESIGN /L8CENSIO90 fRA.E MANB00 i evant ' l raosect rea cr' rot i 9 4 { PL ANT DESIGN , LICENSING, '7,",',",'

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                                               /

SECY Correspondence & Records Branch

   ;   Attached are copies of a Commission meeting transcript and related meeting l   document (s). They are being forwarded for entry on the Daily Accession List and placement in the Public Document Room. No other distribution is requested or l   required.

Meeting

Title:

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                                        \

m k bb @_ 3Q 04~' pevunC Meeting Date: Ic> q( Open Closed Item Description *: Copies - Advanced DCS

                                                                                             '8 to PDR                 Cg 3

1. TRANSCRIPT 1 1

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o' i ORAL STATEMENT SUBMITTED FOR THE RECORD TO \ l UNITED STATES NUCLEAR REGULATORY COMMISSION MODULAR HIGH TEMPERATURE GAS-COOLED REACTOR PROGRAM REVIEW T. E. Northup General Manager Power Reactor Programs GA Technologies Inc.  ! October 9, 1986

.. 4 Mr Chairman and Commissioners: I am T. E. (Gene) Northup, General Manager of GA Tech-nologies Inc.'s Power Reactor Program Group, with responsibility for developing advanced reactors, including our activities on the Modular High Temperature Gas-Cooled Reactor. It is my pleasure to appear before this commission to report on our rapidly escalating progress. The program is well organ-ized. The design is evolving in a very disciplined manner. We are developing a second generation nuclear power system which can satisfy the concerns of the public, the government, the utili-ties, and the investor community about nuclear safety and invest-ment protection. We know where we are going and what we need to do. Now I will try to keep specific on the status of the design without getting too excited, but it is very easy'to get excited about our program. A comprehensive design and licensing basis for the 350 MW(t) Modular HTGR has'been established. Top Level Regulatory Criteria have been developed. All of the participants have concentrated on pulling this design together. It has been defined and it can be defended. A 5-volume Preliminary Safety Information Document was t submitted to NRC at the first of this month -- right on schedule set one year ago. Due to the amount of design work accomplished, we are now prepared to defend this submittal as we interact with NRC. We are after an affirmative licensability statement and Safety Evaluation Report in CY'87. We expect Congress to authorize the $25M required to meet our objectives with NRC in FY'87, including the expansion of design detail.

. 1 All of the top design criteria documents and design descrip-tions have been . issued. There are about 100 of them. They establish and will control the design to the end of the program. These documents include approximately 110 major plant design drawings. Over 43 component top assembly and detail drawings wit,h multiple sheets were issued. The majority of these drawings will be in the PSID. Examples of some have also been included in the handout distributed today. At least 200 computer codes have been adjusted or upgraded for direct use. At least 20 major computer codes have been inde-pendently reviewed and verified by National Laboratories. Design control procedures are in place via the DOE Plant Design Control Office with all participants involved. Many fuel cycles are possible in the MHTGR core but the reference cycle uses uranium enriched to 19.8%, plus thorium as a fertile material, with no reprocessing, meeting all non-prolifer-ation requirements and economical requirements of the utilities. A licensing plan has been submitted and approved by NRC. Twelve topical meetings with the NRC staff and two meetings with ACRS have been held. The most recent meeting with NRC staff was held on september 30, and we expect the pace to pick up as we interact on the PSID in the coming months. A detailed Probabilistic Risk Assessment and Regulatory Technology Development plan will be submitted in the first quar-ter of 1987. Numerous topical meetings are scheduled for FY'87. It is our hope that NRC will maintuin the schedule of activity as planned.

    . .        i J

I think we have been very successful. Four working scale

models have been produced. These models could be made so early in the schedule only because of the progress made in the inter-face design efforts.

A summary of the design and examples of , computer generated 3-D solid interface models are in the separate hand, outs. 4 To date the MHTGR design effort is showing that the goals set by Congress, the DOE and utilities through GCRA can be met with significant margin. The specific goals, set by this Commis-sion in its Regulatory Policy for Advanced Reactors, to provide enhanced margins of safety over current generation LWRs is a rallying point for the participants and a focus of the design. specifically: Passive Safety 1 4 The heart of the MHTGR is its coated particle fuel. Imagine j if you will, a large nuclear power plant containment structure (or your home hot water tank) being shrunk down in size to just cover a piece of uranium fuel the size of a poppy seed. Now imagine 3 more containment structures (tanks) shrunk down around the previous one. Make one more mental picture for me -- make 4 your little shrunken multiple structures (tanks) strong enough to hold the equivalent pressure of 50 containments (25 hot water tanks). You should now have a firm mental picture of the true value of the MHTGR fuel. Every time you see a roll with little i black poppy seeds, I want you to think of this structural feat. I'm a structural engineer and I think it's terrific! 4

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4 our goal has been to develop a power plant. with passive features that will never allow the capacity of the fuel coatings ]. to be violated. We believe we have such a design --'even to the I point where the control room does not-have to be safety grade and the.off-site doses are well below regulatory requirements. Evac-uation or sheltering of anyone outside the 1500 foot site boun-

dary is just not necessary. Natural convection cooling of the core without the use of pumps or valves will be done. In fact, it has already been done in tests conducted on the AVR test reac-tor in Germany over 15 years ago.

How do we know the fuel is so good? Several million TRISO ) coated paticles, to be used in the MHTGR fuel cycle, have been tested in accelerated irradiation tests. over a billion have been irradiated in reactors. This provides an enormous data bank for use in coated paticle design and coating performance. Yes --

                         .these shrunken vessels are designed -- not just made.

Fifty Kg of high quality coated fuel particles for the MHTGR Specification were produced during FY'86 on a full scale pilot production segment at GA Technologies. 1 I The chernobyl accident has drawn attention to reactors using graphite. Large amounts of air and very high temperatures are required to start and sustain graphite oxidation. Here we let the laws of physics help. Due to the silo location of the ) module, the amount of air ingress is so limited that very little I interaction can occur. Even if a fuel element (graphite block

with fuel) was taken right out of the operating core and set

{ outside in unlimited air supply, the graphite will not burn or { exhibit detrimental oxidation. f l i

   .      t Build Quickly - Shoo Fabrication /Short Construction components, including reactor vessel and steam generator, are sized for shop fabrication.                     Projections for equilibrium I

plant - deployment show only 27 months from start of site work to firgt unit power production.

 ;           Mich Availability The concept will meet the 80% availability goal as the small U.S.

Peach Bottom reactor and the German AVR HTGRs have already demonstrated with margin. Investment Protection The plant is designed for decay heat removal by natural convection cooling even with total loss of coolant. As a result, the goal of preventing the loss of the plant investment in the lifetime of a large population of similar plants is attainable. Economy The MHTGR economy looks good. It shows a significant power cost advantage (at least 20%) over coal for the same size of plant. DOE Program Schedule Although . the current DOE schedule recognizes f'unding re-strictions, acceleration would be very possible. The development program does not require large test facilities or infrastructure. l Testing can be paced with the design effort. 1 l 4

u. - - . _ _ . . __. _ . _ . _ _ _ . _ _ _ _ . _ . _ . _ . _ . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _

         . 4

' Present indications are'that the overall cost for the de-1 sign, development, and operation of a single demo plant is about

                  $800 million total and the first demo can be operational by the

] and of 1995, assuming funds estimated for each year are provided. This,would be confirmed by completing the preliminary design work j planned. 1 , International Cooperation International cooperation programs on technology development i for HTGRs are in place with the Federal Republic of Germany and Japan for fuel development, graphite development, fission product j transport technology, metals development, reactor studies and in-core instrumentation technology. The FRG plans for further de-sign development and testing of fuel through 1992 at a cost of about $60 million. The U.S./ Japan exchange is more recent, hav-4 ing been signed in 1986, but similar benefits are expected. l j As a result of privately funded cooperative work, it appears that the Japan Atomic Energy Research Institute may select the ! U.S. prismatic block type fuel element for their Very High Tem-perature Test Reactor program. By the way, the MHTGR is an excellent plant for U.S. export. Succort by Utilities and Vendors GA Technologies is deeply committed to the success of the MMTGR program. We are prepared to apply our energies and avail-able resources to see the program move forward. We do look toward the government participation and support of those areas of first-of-a-kind costs which carry risks that cannot be prudently covered by the private sector. __ ___ ___.__.__-._.__._____-u_._.,_____.

   .       s I trust it is evident from my comments and the materials provided as examples, that substantial progress has been made i

since the program has focused on the 350 MW(t) MHTGR system - a year ago this month. Please continue to help us expand this progress. Thank you for the opportunity to present the status of the MHTGR design before your Commission. I only wiEh that I was more adept at transferring my enthusiasm directly to each of you. i i

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