ML20137G856
| ML20137G856 | |
| Person / Time | |
|---|---|
| Issue date: | 02/21/1997 |
| From: | NRC |
| To: | |
| References | |
| NUDOCS 9704010548 | |
| Download: ML20137G856 (113) | |
Text
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l OfficicI Trcnscript sf Prcceedings o
V NUCLEAR REGULATORY COMMISSION 5
Title:
Industry Presentation on the Use of i
MOX Fuel: Open Technical Meeting Docket Number: (not applicable)
Location: Rockville, Maryland O Date: Friday, February 21,1997 Work Order No.: NRC-1019 Pages 1-112 NEAL R. GROSS AND CO., INC.
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010021 Court Reporters and Transcribers f 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 -
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1 UNITED STATES OF AMERICA !
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2 NUCLEAR REGULATORY COMMISSION lO .
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4 INDUSTRY PRESENTATION ON THE USE OF MOX FUEL !
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5 OPEN TECHNICAL MEETING
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l 7 FRIDAY, i
[ 8 FEBRUARY 21, 1997 i 9 +++++
10 ROCKVILLE, MARYLAND l l
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12 The industry participants met at Two White l 13 Flint Auditorium, 11545 White Flint North, at 8:30 a.m.,
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l 14 Mike Wallace, Chairperson, presiding. j 15 PRESENT.
16 MIKE WALLACE, Commonwealth Edison Company 17 BRIAN COWELL, Oak Ridge National Laboratory l
- 18 SHERRELL GREENE, Oak Ridge National Laboratory 19 STAN RITTERBUSCH, ABB-CE 20 MICHAEL L. TRAVIS, Westinghouse Electric Corp.
I 21 RICHARD D. ANKNEY, Westinghouse Electric Corp.
1 22 CHARLES PAONE, General Electric Nuclear Corp.
4
- 23 JEAN-LUC PROVOST, Electricite de France
] 24 DR. DIETER KREBS, Siemens AG-KWB-B 25 F
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2 1 ALSO PRESENT:
- n. 2 ED LYMAN
\ ) l 3 RALPH MEYER
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4 DAVID EBERT '
5 RALPH ARCHIBAL I 1
6 CHARLIE WILLIS 7 DIANE ARRIGO 8 VANICE PERIN 9
10 i
11 12 l l
13 m
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3 1 A-G-E-N-D-A
,- 2 Acenda Item Page (m) 3 Introduction by David Matthews 4 4 Presentation by Mike Wallace 6 5 History of MOX Use Around the World:
6 sherrell Greene 10 7 Brian Cowell 14 8 MOX Use in ABB-CE Reactors, Stan Ritterbusch 27 9 MOX Use in Westinghouse Reactors:
I 10 Michael L. Travis 39 11 Richard D. Ankney 43 12 MOX Use in General Electric Reactors, Charles Paone 54 13 European PWR Experience, Jean-Luc Provost 77 l
i'v) 14 European BWR Experience, Dr. Dieter Krebs 94 15 16 17 l
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18 19 l
3 20 i 21 22 23 24
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.1 P-R-O-C-E-E-D-I-N-G-S '
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2 (8:53 a.m.) i 3 MR. MATTHEWS: If everyone would please be !
-4 seated, we'd like to begin. i 5- Good morning. My name is David Matthews. I'm 6
the Chief of the Generic Issues and Environmental Projects .
7 Branch.in the Office of Nuclear Reactor Regulation. !
8 Today we are hosting a meeting to discuss the 9 issue of MOX fuel use in commercial nuclear reactors, l 10 I have a few administrative announcements if 11 you'll please bear with me. Smoking is not allowed 12 anywhere inside NRC facilities. So, therefore, if you'd '
13 please refrain from smoking-in either this auditorium or 14 in the outer area there. 'Neither is eating or drinking 15 allowed inside our auditorium.
16 Restrooms are located on this level behind 17 you. '
18 We have with us-today a court reporter who is ,
19 going to transcribe this meeting. If people who are 20 speaking from the floor would please state very clearly 21 their name and their affiliation before making remarks or 22 questioning any of the speakers, it would be appreciated 23 for the benefit of the court reporter.
j 24 Do you have any other requests in that regard?
25 I'm sorry, k nice. Say that again. Is it a NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.
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1 movable microphone?
2 Okay. The expectation is that people would-O 3 walk forward to the usable microphone in the front of the 4 room to ask their questions 5 With that, let us begin. As I mentioned, NRC 6~ is hosting this meeting at the request of NEI for 7 presentations by representatives of the nuclear industry 8 on the use of MOX fuel in nuclear reactors. There is 9 going to be a follow-on meeting in a related area 10 tentatively scheduled for March 26th, which is going to 11 address the fabrication of MOX fuel. It will also be 12 publicly announced, if it hasn't been already, and I 13 believe it will be located in the same place.
rs 1
14 As indicated, this meeting today will be l
\
15 transcribed, and copies of that transcription, when 16 available, will be.placed in our public document room.
3
! 17 Since the publication of the programmatic 4
18 environmental impact statement'in December.1996 and the 19 record of decision in January of 1997 by_ DOE, a lot of 20 interest has been generated concerning the mixed oxide 21 fuel option. The mixed oxide fuel. option is one of the l 4
22 two-choices the Department of Energy has selected for the i
23 disposition of weapons usable fissile materials.
4 i
, 24 After that decision, DOE briefed all of the i
() 25 Commissioners on January 27th, 1997, in a public meeting l
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6 1 here at NRC. In the meantime, NEI briefed each 7- 2 Commissioner and the EDL.
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I 3
As a result of all the generated interest, NEI 4 requested today's meeting to enable them to present the 5 information and experience in the area of MOX fuel use in 6 reactors to the staff and get staff reaction, and of 7 course, as you can see, we have a great deal of industry a and public interest in this subject as well.
9 I'd like to now turn to the first speaker, Mr.
10 Mike Wallace, who is the Senior Vice President of the 11 Commonwealth Edison Company, and he is the Chairman of the f 12 NEI Plutonium Working Group.
/,,
13 MR. WALLACE: Thanks, Dave, and good morning.
l A
's / 14 I, too, would like to welcome you to this open technical 15 meeting, and I appreciate Dave and Vanice helping us to 16 arrange this opportunity for the dialogue to take place 17 between the industry representatives and the NRC.
18 I'd also like to welcome Commissioner Rogers, 19 I see here with us. I appreciate your involvement, sir, 20 and we will be very much looking forward to having a 21 dialogue on issues that are of interest to the NRC based 22 on the discussions that take place from the industry 23 representatives that we have with us here today.
24 Turning to the agenda and a couple of comments i ,
) 25 on how we will proceed in order to assure that this is as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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effective and efficient a meeting as we can make it, given 2 the significant amount of information that we will be O 3 looking to present and yet hoping to dialogue on it a bit, 4
we will handle the meeting in two sections, if you will.
5 The first part of the morning will be a 6
discussion of the history of MOX use from our speakers ,
7 from Oak Ridge National Lab, and I'll introduce them in a 8
bit, and then the next three' speakers representing ABB, 9
Westinghouse, and General Electric will be discussing the 10 results of a study that they conducted for DOE involving 11 the utilization of MOX fuel in light water reactors for 12 which they are presently fuel suppliers.
13 And the essence of their study was to look at t
14 today's reactors from the safety envelope and the 15 technical issues that are involved and identify what the 16 similarities and differences are that we would have to i
17 deal with through the introduction of MOX fuel into !
18 today's U.S. light water reactors. i 19 We'll have all four sets of speakers in the 20 morning go through their presentations before we open it i 21 up for questions, and the reason for that is the synergy l 22 between the discussions is so critical, I think it is best 23 to ask you to hold your questions until we get all done, l -24 and there may be some questions that you wish to address a
() 25 to one individual. There may be others that you would l
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like several of the members to address at once.
2 So the question period will, given where we 3
are in time, will probably start around 10:30, and we may 4 extend some of the late morning times just a little bit to 5 allow ample opportunity for questions.
6 When we come to the question period, since 7
this is an open technical meeting for the NRC staff, our 8 primary focus is on soliciting the questions from the NRC 9 staff and addressing those. Should we exhaust those 10 questions and have time yet available on the agenda, we "11 will receive questions from any members of-the public, 12 which will then be directed to either the industry 13 representatives or the NRC staff, and I'll field those 14 questions as well.
15 At the close of the first half of the meeting, !
16 we'll have a short break. We'll then come back, and the 17 two speakers in the second half of the morning are
- 18. bringing us the European real time experience in 19 pressurized water reactor and boiling water reactor 20 operations. We will have both of those speakers present I 21 their information, and then again open it up generally for 22 questions.
23 And then finally, should there be time prior 24 to 12:45, following the questioning of the European
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1 questions on the whole topic of utilization of MOX in !
9 - 2 light water reactors, and you can draw on the' entire :
1 3s )%-
3 p_anel.
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1 4
One commitment we will hold to is the meeting t j
1 5 will end at 12:45, and because it is a long morning,-I 6 will work to stage the interactions and manage the 7 question periods so that we're able to accomplish that !
i 8- objective.
9 As Dave also mentioned, this is the first of 1
-i 10 two workshops. The second workshop -- I'm sorry --~open i i
11 technical meeting. The second open technical meeting is 12 going to be on March 26th, and it will fundamentally deal 13 with fabrication of MOX fuel. As such, that topic is not 14 one that is relevant for this meeting. This meeting is 1
15 focused on light water reactor operations with MOX fuel.
16 What I'd like to do then is. introduce our i 17 first speaker of the morning, Sherrell Greene. 'Sherrell 18 is Manager of Fissile Materials Disposition Program at Oak 19 Ridge, and for the last year, he has led the reactor-based 20 fissile materials disposition activity at Oak Ridge.. He 21 has responsibility for the technical and economic analysis 22 of reactor options and over 20 years' experience in the 23 field.
24 In the interest of time, Felix, what I'm not
() 25 going to go into is a discussion of the mission statement, NEAL R. GROSS COURT REPORTERS AND TRANSCRSERS 1323 RHODE ISLAND AVE., N W.
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1 I which is :in your handout ~, and the objectives of the !
,, 2 meeting, which is in your handout, and the listing of the 3
participants of the NEI working group, all of which are in 4 your handout, but I will make perhaps one point on the 5 working group before Sherrell begins.
6 And that is the working group now censists of 7 41 members, 11 utilities and virtually all of our fuel 8
fabricators and those who are involved in Mox fabrication 9 around the world, and it is a committee of the whole, 10 which means that we invite any NEI member to join the 11 committee if they have an interest in the disposition of 12 weapons grade plutonium through the utilization of MOX 13 fuel.
14 Those are the members of the committee. The 15 mission is contained in the handout. The objectives of 16 the meeting are focused in the way that I outlined just a 17 minute ago, and now I will, indeed, introduce Sherrell.
18 Thank you.
19 MR GREENE: Thank you, Mike. '
20 For the last three years or so as part of the l
21 Fissile Materials Disposition Program, Oak Ridge National l
- 22 Lab has had the role of developing concepts and performing 23 technical and economic assessments of various reactor l 24 based plutonium disposition options. In a few moments
() 25 we're going to present to you an overview of the history NEAL R. GROSS i COURT REPORTERS AND TRANSCRSERS I 1323 RHODE ISLAND AVE., N W. ;
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11 1 of MOX utilization in light water reactors around the
,-s 2 world, and what I wanted to do first is to provide a
'] 3i i
little bit of context for that discussion and for the 4 remaining discussions of the day.
5 It's very important to realize that the FMDP 6 MOX technology that's being discussed now is an 7 evolutionary, that is, evolutionary, development of 8 existing commercial LWR MOX technology. Indeed, LWR 9 technology, as we all know, is in use around the world.
10 What is surprising to some people is that the 11 history of LWR MOX technology dates back and begins 12 actually with the use of weapons grade and near weapons 13 grade material for the manufacturing of many of the (m )
-k/ 14 initial test assemblies.
15 And so there is experience both in the U.S.
16 and Europe in the utilization of weapons grade and near 17 weapons grade material in MOX fuel used in light water 18 i reactors, and that's indicated by this little sphere at j 19 the top of this viewgraph.
l 20 Currently, as we all know, commercial LWR MOX 21 is used in Europe. It has a very good history of 22 performance. Our vision for the FMDP MOX technology is 23 that we would begin simply by utilizing fuel management 24 schemes that are very similar to those that are in use gm,
) 25 today in commercial reactors burning MOX. However, it NEAL R. GROSS ,
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would be done, of course, with our weapons grade material, f-~s 1
2 which is signified with a little sphere there in the lower 3 right.
4 Eventually, if the need arises and if the 5 technology permits, we would evolve, again, if the need 6 arises, to full core MOX, but only if the need arises, and 7 I should say also that the terms " partial core Mox" and 8 " full core MOX" tend to be loosely defined and used in 9 different ways by different individuals and organizations.
10 Let me say here my intent is that by partial core MOX what 11 we mean is those fuel strategies, those fuel designs and 12 core management strategies that allows one to avoid the 13 use of integral, burnable poisons in the fuel. We believe l
)
\/ 14 that is a technical issue which can be addressed in a 15 rather straightforward fashion, but probably is not 16 worthwhile to do so unless one really needs that kind of 17 disposition capability.
18 Next slide.
19 Now, how is the MOX technology that we .
20 envision for this program different than that that is in 21 use around the world today? There are two major -- the 22 two major differences, in the sense that they are 23 distinct, not necessarily that significant
,l 1
24 technologically, but they are distinctive -- are, of l
,- m (x_- ) 25 course, weapons grade material has a fissile content. We I 1
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.13 l' have more Plutonium'239. We have less PU-240, and we have l I .
2 less Americium 241 than is present in the commercial MOX 1
3 in use today around the world, manufactured from recycled 4 plutonium. ,
l
?-
5 So the isotopics is one difference, and that 6 is an inescapable difference. I 7
The other difference is that the baseline 8 technology for the-program now involves dry plutonium
} 9 conversion and purification processes. They offer some !
- 10 environmental advantages that are very desirable, and we 11 would like to pursue those.
12 However, there are two artifacts'of the use of 13 these dry conversion technologies. One is that there's a O.
14 possibility that-trace element compositions might be i l
different'in the plutonium and in the plutonium oxide, and 15 l
16 there will be a different_ powder morphology. The
'17 micromorphology of the powder will be different. l
.18 There are a variety of R&D activities underway l
.19 to demonstrate the acceptability of this. We believe that !
l 20 will be done in a straightforward fashion. However, it is '
21 worthy of note that if for some reason that did not occur, 22 that we, of course, have the back-up of falling back on I i
23' current industrial practice, which we would still employ j 24 the dry plutonium conversion for the pit disassembly. We l
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14 1 precipitation, and we would be right back with the same s
_ 2 material that is in use today, and that, of course, would e
\/ '
3' give us the standard powder morphology and give us a very 4
- mprehensive plutonium purification capability.
5 So this is the context in which you should 6
view the discussions that you'll hear now and later today.
7 These are the distinctives that we are discussing. The 8 isotopic issue, short of isotopic blending, will be there.
9 There are straightforward technical approaches to the 10 other two issues that I've laid out there, and I would ask 11 you to keep this in mind.
12 I'll now introduce Brian Cowell from Oak Ridge 13 National Laboratory. Brian is our lead MOX qualification 7
k_) 14 engineer. and Brian is going to present an overview of the 15 history, the worldwide history, of the use of MOX fuel in 16 light water reactors.
17 Brian.
18 MR. COWELL: I'm going to present a summary of i 19 the work I've been doing for approximately two years to l
l 20 recover the historical database for MOX fuel use both 21 domestically and abroad.
i l
l 22 As Sherrell said, my name is Brian Cowell.
23 Next slide, please.
24 This presentation summarizes the history of
[) 25 white water reactor mixed uranium-plutonium oxide or MOX v
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15 1 fuel.
This white water reactor MOX is distinct from the 2 uranium-plutonium oxide liquid metal reactor fuel that O 3' scme people refer to as MOX, but the more common l
4 interpretation of MOX is for light water reactor MOX only.
5 Some differences are a lower plutonium content 6 in the light water fuel, different cladding typically, a !
7- higher fuel density, a lower center line operating 8 temperature, and a difference in the oxygen-to-metal 9 ratio.
10 I don't intend to discuss further any of the 11 plutonium fuels work for the liquid metal reactor.
12 The rest of the presentation is organized by 13 either country or region. First I was to describe the O 14 domestic experience base, followed by a discussion of the 1 15 European' development, and then two special other 16 countries,.the. Japanese development and Russian 17 development, and then a brief summary.
)
18 MOX fuel R&D started in the '50s before cold-
-19 pressed and centered pelletized fuel became the de facto !
20 standard. Early work was performed under the Atomic 21 Energy Commission sponsorship.
22 Some of the programs were the Plutonium i j 23 Utilization Program at-Hanford, which began with i
l 24 analytical work, critical experiments, and some test I
(( )
r 25 reactor irradiations in the materials test reactor and l NEAL R. GROSS l COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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- 3 l 1 engineering test reactor.
2 A special purpose reactor, the~ plutonium 3 recycle test reactor, was built at Hanford in which over
- n 4 4,000 rods were irradiated in the '60s.
5 Another program was the Saxton Program, which 6 was'AEC sponsored in a subcontract to Westinghouse. Over 7
600 rods were fabricated and irradiated in the Saxton Core 8 II and III.
9 A Boiling Water Reactor Program was sponsored 10 by the AEC~in a joint program with General Electric and 1
11 Euratom. Testing was performed on BWR rods in the 12 Valicitos BWR and in VR-2, which is a Belgian reactor, i
13 Another experimental program was VIPAC fuel 1
l O 14 tests performed in the experimental BWR at Argonne, over 15 1,200 rods.
16 Many of these early tests utilized VIPAC, 17 Sphere-Pac,. swage compacted, hot pressed pellet, and/or 18 annular fuels that have only limited applicability to l 19 modern cold-pressed and centered solid pellets.
1
.20 After the early development work, by the late 21 1960s commercial demonstration programs under private !
( i 22 sponsorship developed. Two of the more important of these i 23 were in joint program with Edison Electric Institute. !
l l 24' The firs' ., EEI-Westinghouse Plutonium Recycle 25 Demonstration Program, began in 1967 to investigate MOX J
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1 17 1 fuel use in PWRs. It culminated in the insertion of four
(
73
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2 LTAs for a total of 720 rods in San Onofre I in 1970, and
' \
3 they were kept in the core for two cycles.
1 4 A parallel program between EEI and General 5 Electric, the Plutonium Utilization in BWRs Program, also 6 started in 1967 and continued the early AEC G.E.-Euratom I 7
work, and the rods listed there were all tested as part of 8 that program. It included demonstrations in Dresden, Big I 9 Rock Point, and Quad Cities.
10 Other privately sponsored programs existed.
11 Two of the primary programs were with Exxon Nuclear and 12 with Gulf United Nuclear. The testing was performed in 13 the reactors listed.
,/
w/ 14 In all of these commercial demonstrations, the 15 MOX fuel performance that was experienced was roughly 16 equivalent to contemporary LEU fuel. There were some 17 difficulties with a small number of the rods, but these 18 were difficulties that were also being experienced with 19 LEU fuel at the time, and there was no performance that 20 was perceived to be problematic that was attributable to 21 the MOX fuel itself.
22 In the early 1970s, the AEC initiated 23 preparation of the generic environmental statement on the l
24 use of recycled plutoniun, and mixed oxide fuel in light
( ) 25 water cooled reactors, or GESMO. The AEC determined that NEAL R. GFU3SS l
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18 1 adoption of rules governing wide scale MOX use constituted 2' a' major federal action having the potential to affect 3' significantly the quality of the human environment, and 4 this required an EIS in accordance with NEPA, Section 5 102.2(c).
6 The resulting GESMO review covered.all aspects
-7 of plutonium recycle, not just reactor' irradiation.. The 8
first draft of GESMO was published in 1974 by.the AEC, put 9' out for public comment. A final version incorporating the 10 public comment was published as NUREG-0002 in 1976 by the 11 NRC.
12 A safeguard supplement to GESMO was prepared 13-- at the request of the President's Council on' Environmental D
\'- 14 Quality, and public hearings began in~1976, following the 15 publication of that NUREG.
16- GESMO findings, in gene'ral, were favorable 17 towards MOX utilization. I've taken two of the findings 18 from GESMO. The safety of reactors and fuel cycle 19 facilities is not affected significantly by recycle of 20 fissile materials, and the GESMO assessment showed that 21 the potential hazards to the public for the model mixed
'22 oxide fueled light water reactor remain relatively
.23 unchanged by the_ substitution of mixed oxide fuel 24 assemblies for uranium fuel assemblies for both normal and
() ~ _2 5 accident conditions.
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19
- 1 By this time, in the mid-1970s, the domestic I 2- MOX fuel industry was ready for commercialization.
,jJ k-
\ MOX l 3
fuel utilization at that point was limited by plutonium ,
, t 4 availability, government delays in granting license ,
5 requests, which was required to complete GESMO, and l r
6 finally, by the MOX fuel fabrication capacity, which was i
7 limited at that time to lab and/or pilot' facilities at 8 several locations around the country.
2 I
4 9
The fuel vendors in the mid '70s believed that i' I 10 the LTA and the limited reload experience provided a i
11 sufficient database for commercial implementation of MOX 12 fuel. Domestically, approximately 1,400 rods of both BWR l
13 and PWR had been tested, plus some additional rods which I [
(f 14 don't include in the commercial database.
15 Additionally, foreign work performed by the 16 domestic vendors occurred in a number of European 17 reactors, which I'll' discuss in just a moment. l 18 However, at this time, U.S. nonproliferation 19 policy halted all of the MOX development, and on April j 20 7th, 1977, an executive order on the Statement on Nuclear 21- Power Policy and a subsequent clarification letter J
22 requested termination of the GESMO proceedings, 23 termination of all proceedings on pending and future l l
l 24 plutonium recycled related license applications, except
() 25 for those licenses for experimental purposes only, and NEAL R. GROSS l COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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i 20 1 finally, publication of the draft safeguard supplement to 2 GESMO.
4 3
i All of the ongoing R&D was either halted i
4 immediately, converted to research on alternative fuel I 5 cycles, or completed without fanfare, and several of the 6
ongoing R&D programs were entirely sponsored privately at 7
that point, and those for the most part were completed.
I i 8 The four assembly irradiation in Ginna was the 4
9 only significant activity that was allowed domestically 10 following the policy announcement in 1977, and this 11 irradiation was of some rods produced by Westinghouse in 12 the earlier.-- prior to this announcement, that had been '
13 stored until 1980, and at that point, the utility 14 requested to load these bundles, and this was allowed,'in 15 part, because it provided for disposition of nearly half 16 of the separated commercial plutonium remaining in the 17 U.S.
18 This slide summarizes the domestic experience 19 base. It does include the Saxton rods, but does not 20 include the EBWR rods. It shows the distribution of when 21 the rods were initially inserted into the reactor, and 22 this is the summary of approximately over 3,000 rods.
23 European MOX fuel research and development 24 followed a similar path. Some of the early national
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I 1 European countries. These early programs performed i I: 2 analytical studies and critical experiments and test rod '
E~
4 3 irradiations, and these programs were undertaken in l 4' Belgium, France, Germany, the U.K., and Italy.
I 5 The Belgian program is of note because it led 6 to-the world's first PWR MOX irradiation in the BR-3 7 reactor in 1963.
8 And during the 1970s, European technology 9 advanced through LTA testing to commercial recycle.
10 Almost 15,000 MOX fuel rods were irradiated in various 1
11 European commercial light water reactors as part of either
- .12 LTAs or recycled reloads during the 1970s, and many I
13 countries had active MOX programs at this time.
lO j b 14 Germany was irradiating in a'PWR and two BWRs; l
)
! 15 France in the Chooz BWR; Italy in both the BWR and a PWR; -
16 the Netherlands, Sweden, and Belgium also had irradiation-17 programs.
18 The utilization of MOX fuel in. Europe at this I
19 time was limited primarily by the shortage of plutonium
)
20 that was caused by delays in the reprocessing capacity.
21 The fuel performance that was experienced with this mid- i l
22 '70s fuel was for the most part satisfactory and, again, 1
23- equal to contemporary LEU fuel, the same result that was !
24 found with the U.S. testing. i 25 However, by 1980, liquid metal reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W. .
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22 1 implementation was expected to be rapid, and this 7
s 2 implementation was expected to eliminate the need for MoX
( ' '
/
3 recycle in the light water reactors. Therefore, many of 4 the national programs switched their focus frou light 5 water Mox to liquid metal reactor plutoniun fuel.
6 Only two programs continued MOX development in 7 earnest, and these were the KWU or Alkem Program and also 1
8 the Belgonucleaire Program.
1 9 The improvements to MOX fuel at this time were 10 focused primarily on improving the solubility of MoX for 1 11 the recycle industry in Europe, and this was done through 12 increasing the homogeneity of the fuel.
13 And during this type MOX recycle did continue
(. l 14 on a limited scale in the German reactors listed and also 15 in Switzerland.
16 By the mid-1980s, the picture had changed in 17 that the LMR industry was not expected to expand as 18 rapidly as earlier predicted, and in 1985, EdF decided to 19 recycle their plutonium in their 900 megawatt units. The 20 first EdF MOX fuel was loaded in 1987, and EdF MOX use has 21 continued to expand since then, currently in seven of 22 their 900 megawatt units. Another nine are licensed, and 23 the rest of the 900 megawatt fleet is considered j 24 technically feasible for MOX fuel utilization.
?
( ) 25 Currently MOX utilization also continues in
- s. -
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23 1
other European countries, including Switzerland, Belgium, 2 and Germany.
O 3 The European experience has proven the 4 feasibility of MOX-fuel utilization. Approximately 5 300,000 MOX fuel rods have been irradiated. The self-6 generated recycle or partial core loads -- these'are 7 sometimes used synonymously -- which is approximately 30-8 percent of the rods in the core is MOX fuel, can be 9 accommocated with little impact on the reactor.
10 The in-pile performance of these 300,000 MOX 11' rods has equaled or in some dases exceeded that of the 12 equivalent LEU fuel. produced at the time, and the vendors 13 claim that the burn-up limits for MOX fuel should be 14 identical to the LEU limits based'on their testing.
15 Now, I would like to switch to thefJapanese 16 MOX fuel history,.which has focused primarily on the Fugen 17 ATR utilization up until very recently. In Fugen, the 18 advanced thermal reactor is a heavy water moderated 19 boiling white water cooled design that was developed in 20 ; Japan exclusively.
21 Fugen was designed around a plutonium fuel 22 cycle, and since its initial operation in 1979, Fugen has 23 burned over 600 MOX assemblies containing over 18,000 MOX 24 fuel rods, all produced by PNC.
() 25 Recently, MOX utilization in Japan has shifted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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t 24 1 away from the ATR design towards the LWR fleet. Two 2 island designed MOX bundles were loaded into a BWR in
[ ,\
3 o
1366, and four all MOX LTAs were loaded into a PWR in n
4 lias, and future MOX utilization is planned in a new fleet 5 of advanced boiling water reactors being constructed 6 currently.
7 Now a brief mention of the Russian MOX fuel 8 utilization, which to this point has been limited to a 9 very few test rods. All the plutonium in Russia 10 previously has been dedicated to their liquid metal 11 reactor programs, and in fact, they have shortages in 12 those programs and were usir.g uranium fuels in some of 13 their fast reactors.
r I
-/ 14 The fast reactor fabrication processes that i
15 were used may be suitable for LWR MOX with some minor ;
1 l
16 modification. However, only a few LWR .:.OX test rods have '
17 been produced so far, and these were tested in the MIR l
18 test reactor in what is termed by the Russians life 19 testing. A plan is in place to insert three MOX LTAs into 20 the Balakova-4 VVER within the next five years.
21 Finally, I'd like to summarize a statement on 22 the isotopics of the plutonium that's been utilized to 23 date. Much of the early MOX fuel testing included near 24 weapons grade plutonium because reactor grade plutonium at 25 the time was simply not available. Much of the early AEC (w.)
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1 25 1 work including some of the commercial irradiations l
f
\
2 utilized what was termed AEC plutonium, which in fact was
> 3 4 0 percent fissile. This included some of the rods in Big 4 Rock Point, Quad City, and Quad Cities, and the San Onofre 5 reds were just slightly degraded and were 86 percent 6 fissile.
7 Additional material that was used in these 8 early tests was obtained from Dresden recycle, and this 9 material was from very low burn-up fuel and was 10 approximately 80 percent fissile.
11 It's known that some of the overseas 12 irradiations also included greater than 90 percent fissile 13 plutonium, and finally, many of the critical experiments,
,,3
(,) _
14 both domestically and abroad, utilized this near weapons 15 grade material.
16 The Plutonium Utilization Program experiments 17 tested a range of plutonium fissile contents up to 92 18 percent fissile.
19 In Japan the deuterium critical assembly was 20 utilized in support of the Fugen ATR, and in this test 21 greater than 2,200 rods were produced and tested utilizing 22 91 percent fissile plutonium.
23 In summary, MOX fuel utilization is supported
! 24 by a long history of success. Partial core loads of s
25 reactor grade MOX fuel have been ope?;ated for tens of (v)
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26 1 reactor years both domestically and abroad.
I
,s
)
2 The overall fuel performance of this MOX fuel
~'
3 has been similar to LEU fuel of the time. With the 4 current homogeneous cold-pressed and centered MOX fuel, no 5 failures attributable to the unique properties of the MOX 6 have been experienced. There have been some limited fuel 7 failures, but none of this was attributable to the 8 uniqueness of the MOX.
l 9 MOX burn-up limits are currently lower in 10 France than the LEU burn-up limits, but EDF is actually I
11 pursuing equal burn-up limits for the two fuels, and their l 12 fuel fabricators believe this is supported by the 13 database.
/'N
( )
\~ / 14 The isotopics, as I just mentioned, of the l 15 surplus weapons grade plutonium are not significantly 16 different from the early plutonium that was used in the 17 testing.
18 And finally, plutonium that's generated in 19 situ in low enriched uranium fuel originates as weapons 20 grade plutonium, and so it's important to realize that in 21 low burn-up there is little difference in that fuel and in 22 the fuel we're discussing in terms of isotopics.
23 That summarizes what I've done.
24 Thank you.
73 (v ) 25 MR. WALLACE: Thanks, Brian and Sherrell, for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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27 1 a good overview. I think that well sets the stage for the
'2 rest of the presentations that are going to take place in
,O 3 the course of the day.
4 The next three speakers all are going to be 5 talking about the experiences that they had in the DOE 6 funded activity, to review the potential for MOX fuel 7 utilization in currently licensed light water reactors.
8 The first speaker is Stan Ritterbusch from 9 ABB-CE. Stan has over 26 years' experience with ABB-CE, 10 and that includes both as a fuel designer and in 11 performing safety analysis for ABB and SSS systems.
12 In addition, he's been a project manager in l 13 the past and is currently their Manager of Standard Plant 14 Licensing.
15 Stan.
16 MR. RITTERBUSCH: The analysis we performed 17 was for our, System 80 reactors currently operating today.
18 These reactors were designed for MOX fuel usage in the 19 1970s. So today, for this analysis, not very much has 20 changed.
21 This slide shows a summary of the results. We 22 determined the throughput and the emission time for 23 different options. The first option is shown in the left-24 hand column. That is a MOX fuel design without Erbia l) 25 poison mixed'in with the Mox fuel. For that design, we NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE.. N.W.
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28 :
1 have a throughput of approximately one ton of plutonium l 1
2 per reactor per year. l C
3 so if we have a 50 ton supply of plutonium 4 with three reactors operating, tae mission time would be s' 17 and a hnif years. Similarly, for a supply of 32 and a j 6 half tons of plutonium, the mission time would be 11 and.a !
7 half years.
8 One of the basis ground rules for this 9 analysis was that we would not use Erbia in the MOX fuel 10 pen, and this is a manufacturing consideration which other 11 speakers can address later.
12 However, in order to find out what this option 13 or this ground rule is costing us in terms of throughput, i f 14 we did'the analysis, and we found out that we could l
15 process about 1.2 tons per year per reactor in somewhat i
j 16 shorter mission times that you can see in the lower right- l 17 hand corner. It is our view that this small improvement 18 in mission times means we chose the right ground rule.
19 Other ground rules have to do with changes or, 'I 20 shall we say, lack of changes to the current design and !
21 operation. The first rule was to use existing fuel and 22 core design approaches. Those are analysis methods.
i i
23 The second is to match the current reload !
s 24 cycle for uranium oxide cores currently used by the 5' 25 utilities, and of course, the idea.here is minimize the
- . NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISt.AND AVE., N W.
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29 j i
i disruption of plant operations. ;
t 2
Similarly, we wished to match the cycle length j i
3 for existing reload cores, and also. minimize plant i 1
4 hardware and operations changes.
i 5 With respect to licensing, we also wished to i 6
remain to the extent practical within the current plant l t
7 technical specifications, and this helps minimize the !
a licensing risk, and of course, for economy we wished to !
9 maximize the energy extracted. !
10 If the only criteria were to meet the spent {
i s
i 11 fuel criteria for plutonium, and that is you burn it !
12 enough to mix it with fission products and make it l l
13 unmanageable, the mission times would be much shorter, but l
) 14 we wished to have an economical operation as well.
.15 Next slide, please.
]
i 16 As.I mentioned, the System 80 reactors were 17 designed in the '70s to handle MOX fuel. The most 18 significant features are a large control element assembly 19 worth relative to other designs. We have spare CEA 20 locations, should they be required.
-21 With respect to plant support systems, we have 22 additional delay heat removal capability. The plant 23 systems can also accommodate higher required soluble boron 24 needs.
25- We have a revised and improved reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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1 i
30 1 1 internals design that enables us to put more. control 2 element assemblies into the core, and we also reviewed the 1 3 fuel handling and storage equipment.
4 The System 80 reactors are operating at a 5 power level of 3876 megawatts. They are relatively new 6 units. The dates of the initial operation of the three l-7 units are shown on the second line, and one feature, 8 since there are three units at one plant site, this would
! 9 enable the mission to be accomplished at just one site, 10 and that is a consideration with respect to 11 transportation.
12 Other design features are the use of -- well, i
13 each reactor uses 241 fuel assemblies in the reactor core.
14 Each fuel assembly has a 16-by-16 fuel pin array. There 15 are 89 control element assemblies of various sizes that 16 give us flexibility in how we manage the fuel cycle. You 17 see we have 12-finger control element assemblies, four-18 finger assemblies, and then the four-finger part length. ,
! I 19 We do not expect to use the' spare CEA locations,.but they 20 are there. !
21 Next slide, i i I 22 At the top of this slide you see a cross- l 23 section of a typical fuel assembly. The large holes are i
- I
! 24 for'the insertion of control elements. Each fuel assembly l f( -25 in the reactor core has these five holes, which gives us NEAL R. GROSS l COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.
I (202) 234-4433 WASHINGTON, D C. 20005-3701 (202) 234 4433 !
l- !
i
31 i
i flexibility of inserting the control elements into any {
-l
! 2 fuel assembly which maximizes the fuel shuffling and O 3 flexibility for fuel cycle management.
'4 This slide shows a cross-section of the !
5 reactor core. It may be a little hard to read. The l 6 reason for that is you see a large number of control i 7 element assemblies. If you did the counting, you.would !
t i
8 see control element assemblies being inserted into j l
1 9 something like 85 percent of the fuel assemblies in the i
10 core. This gives us good power control and a high control i 11 element shutdown worth. I l
12 Next slide, please. I 13 This shows a vertical cross-section of :he i 14 reactor vessel. The most unique feature is shown at the 15 middle of the diagram where we see a number of control 16 element guide tubes or shrouds. The control elements go 17 through the middle of these shrouds as they are inserted 18 into the reactor core.
19 The unique feature is that there is one shroud i
20 for each control element or control finger. That is 21 approximately 1,200 total. This large number of CEAs, of 22 control shrouds is a critical feature in being able to l 23 insert CEA assemblies into almost all fuel assemblies in l l
y 24. the core.
I-() 25 I will now summarize the analysis for this l
i NEAL R. GROSS l' l COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N W.
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32 i i particular MOX design. The general process for designing i
2 a MOX' reactor core occurs in two steps.
!O 3 to design a number of different fuel assembly types The first step is I
4 irrespective of how they will be used in the core. !
i 5 This slide shows.two different types with i i
6 different combinations of the Erbia fuel poison rod pins. :
7 Those are the very dark squares. The other squares are i
8 MOX fuel pins. '
i 9 There is some guesswork done in selecting 10 these designs. It is guesswork based on previous 11 analyses. For this particular work, we designed ten such 12 fuel assemblies.
13 The second part of the design process is to 14 select different combinations of these fuel assemblies and 15 assemble them into a complete core, and then do the 16 appropriate analysis.
17 This slide shows some of the parameters. Some 18 of them may be a little hard to read, but we won't go 19 through all of them. It can be observed, however, that 20 many of the parameters are the same. I guess from a 21 designer's point of view,.this is not very exciting, but 22 it is good because it minimizes the change to the fuel i
-23 cycle and utility operations.
l 24 In the middle of the list, we will see one 25 difference, however, and that is in the design of the f()
NEAL R. GROSS-COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 1 specific fuel pins. For the typical cycle, we have a
,S 2 maximum Uranium 235 enrichment of about four and a half b 3 npercent. For this particular analysis, we limited the H,
4 percentage of plutonium in the MOX rod to six and one-half 5 percent.
6 The Erbia poison rods use Erbia oxide mixed in 7 a matrix of natural uranium. This is essentially 8 unchanged from current designs.
9 This slide shows four of the many cases that 10 were analyzed to come up with different core designs. We l 11 look at different side reload batches. The first case l 12 shows a reload batch of 88 fuel assemblies. The second
_ 13 case shows a reload batch of 81 fuel assemblies. I'm ;
')
14 going to talk about the second case because that is the 15 case which gave us the best result with respect to mission 16 time.
17 For this particular case, we analyzed 25 fuel 18 assemblies that had 212 MOX fuel rods in each assembly.
19 We also used 56 assemblies with 180 MOX rods,.and the 20 other columns indicate -- if we move two columns over, we 21 would see then enrichments in the range of from 4.7 22 percent to six percent.
23 Next slide.
, 24 This slide shows some of the neutronics rx
() 25 parameters which are calculated as a result of using the NEAL R. GROSS CoVRT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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34
- core design that I just mentioned.
' 1 The first column shows I a
2 typical uranium oxide reload cores, and if we look at Case
.( i 3 2, which is the third column over, we will see radial I 4 power peaking factors which are about five percent higher
- 5 than the typical reload core. We also see a somewhat i
6 higher enriched boron concentration,.and of course, the i (
- 7 reason is that the boron is worth less. So there has to be more of it to provide the appropriate reactivity worth.
~
8_
f 9 If we go down looking at the moderator 10- temperature coefficient and the fuel temperature .
11 coefficients, we see that they are somewhat more negative.
12 This is a benefit for certain plant accidents. As it
, 13 turns out, it is a significant benefit for the limiting of J s_) 14 plant accident, which is a control rod ejection event. ]
4 I 15 At the bottom, we see that the MOX fuel design l 16 has a lower effective delayed neutron fraction and a lower 17 neutron lifetime. For the control rod ejection event, j 18 which is a large reactivity insertion event, this is not ,
19 so good. It hurts us a bit. !
20 However, the mitigating factors are that the 21 ejected rod is worth less, and the fuel temperature i 22 coefficient which helps mitigate the event is improved.
23 So, in fact, our preliminary analysis indicates that the i
24 existing accident analysis bounds the accident analysis '
i
() 25 for MOX.
NEAL R. GROSS i COURT REPORTERS ANO TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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I 35 1 These. data show the control element assembly
(
2 worth for different combinations. The conclusion here, i
3 this is the evidence that shows that the control element I 4 assemblies are worth slightly less for a MOX core, and j
5 this, of course, is basically the result of the hardened l 6 neutron spectrum that occurs when you have more
! 7 significant amounts of plutonium in the reactor core.
8 This slide shows the results of the four 9 cases, and you will see that Case 2 has the results that I 10 indicated on the very first slide.
l 11 We lookno ary carefully at those parameters l
12 which we thought might affect the accident analysis. As I l 13 indicated the two parameters of most concern are the t
\s- 14 effective delayed neutron fraction and prompt neutron i
H 15 lifetime. These have an impact on the control rod I 16 ejection event.
l 17 However, as I indicated, we expect acceptable )
18 results, and we would expect that when we do the final
! 19 confirmatory analysis that those results will be very J ;
i 20 close to current results or possibly less severe than 21 current results. 1 22 With respect to licensing of a MOX fuel rod l l
l 23 for this particular reactor design, we considered all .
)
24 aspects, including the impact on plant systems and
() 25 equipment, the impact on core design safety analyses and L NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 i i
i the technical specifications. We looked at the impact on '
.. 2- the analytical 1models and fuel qualification data, and I
(
3 also the impact on the environmental report. I 4 It'is our belief that because not very much I 5 has changed.with this reactor design by way~of design
~
6 parameters and because the reactor was designed that way, 7 and because we used many of the reactor core parametera i
8 that were similar to or bounded by current mixed oxide !
i 9 reload cores, that we would expect the NRC' staff to come (
i
'10 to a finding of.no significant hazards. !
l 11 In our terms, thit means not very much has !
l t .
12 changed. We do expect a lot of questioning, and we would !
l 13 have to go through a lot of detailed formal analyses, but l
' (~ l 14 our judgment today is that not very much would change. :
15 I now have several slides quickly stepping l g
16 through those licensing elements that I just mentioned. I 17- The first is the impact on plant syctems and equipment. '
L '
18 We could potentially replace the 13 part length control 19 element assemblies with full length, full strength control j .
20 element assemblies. We do not have to do that, but it
-1 l, '
l 21 remains an option.
! ~
! 22 We will use higher boron concentrations, i
, 23 However, all plant systems are already designed to handle 24 that.
25 The other plant systems which are already NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.
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_.. , .~. _ _ . . . _ _ _ , . ._ _ .
37 1 designed for the MOX core are the shutdown cooling system. !
.- 2 It has adequate capacity. We also have adequate spent t
3 cool fuel cooling capacity. The neutron effluence on the l
4 reactor vessel doesn't change very much, and our rad waste l 5 systems and fuel storage and handling systems are capable.
6 You've heard now several times and you won't 7 be surprised at the first bullet. The fuel design hasn't 8 changed very much, and we do not have to change the part 9 length, part strength rods, but we could if we needed the 10 extra shutdown worth, and the increased boron 11 concentration provides the additional reactivity, negative 12 reactivity worth, that we needed.
_ 13 We looked at the control rod ejection event, i '
\- 14 which is the most severe reactivity insertion event, and 15 we think that that is covered. So we would expect that i 16 our current safety analysis bounds the cafety analysis of 17 ene MOX core. of course, this would all have to be l 18 confirmed in some detail.
19 The next slide, please.
l 20 It gets a little more serious when we look at i
! 21 the codes that we use for fuel design and fuel performance 22 analyses. We have to look very carefully at the l
23 qualification of those codes and need to insure that they 24 are benchmarked against the most current MOX data.
( ,) 25 We also need to look at the design of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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j' i i :
38 !
- 1 particular MOX. fuel rods that we would expect to use and 1
i
{
2 either insure that they're covered by existing programs or 4
j s
3 go through a special program to insure that the design we
, 4- would use is qualified. f i
, 5 With respect to the impacts on the ;
i environmental report, we don't think that there are 6 i I
{- 7 significant impacts because the number of fuel assemblies i
~
t 8 which would be produced is the same. The power generation is the same.
- j. 9 The fuel design is the same, and the number l e ,
! 1 l- 10 of assemblies would be the same. j
{ :
1 11 -And moreover, for the particular assemblies, ;
j 12 the features of those assemblies are very much the same. !
!. 13 I would like to point out that and emphasize that the f'
J 14- current reload cores for -- shall we say spent fuel from i 15 current reload cores using mixed oxide fuel also contain !
i
. i 16 plutonium. So the fission product mix we do not believe I e
d
[ 17 is significantly different between the uranium oxide and i 5 ,
18 the MOX, spent MOX fuel rod. !
5 I j 19 Last slide, please. ;
i i 20 This' summarizes the licensing aspects. No i
21 plant modifications would be required for this program at l j 22 the current System 80 plants. We could replace the 1
j~ 23 control rods, and we would increase the boron 2
j 24' concentration. We do not think there is any measurable 25 impact on the environmental statement.
p i
NEAL R. GROSS COURT REPORTERS AND TRANSCRl8ERS
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i r i 39 l 1 The most significant issues, we think, are -
2' generic policy questions on the use of MOX fuel in ,.
3 eperating reactors. For our particular design, we are !
j 4 most concerned about benchmarking the analytical codes j 5 against the most current data and the qualification of the
?
\
6 particular fuel design. !
- 7 MR. WALLACE: Thanks, Stan, f j j
, 8 Our next two speakers are from Wastinghouse. $
l l
- 9 First will be Mike Travis. Mike is Manager of Nuclear l i
e 10 Development Programs at Westinghouse and has 28 years' i
11 experience in the nuclear industry. He's worked on the
]
. 1 l 12 liquid metal reactors, light water reactors, and other 3 13 specialty designs. Perhaps most notably in the.'70s and i ),
s
)
3- 14- '80s he designed MOX fuel for liquid metal reactors.
15 And second, Rich Ankney has over 20 years in 16 the nuclear industry, primarily in the area of reactor 17 core design, and he is currently an advanced technical 18 engineer in the Core Engineering Department of 19 Westinghouse, responsible for advanced reactor core design
'20 and analysis. In the past several years, he's been 21 analyzing MOX fuel usage in Westinghouse light water i
22 reactors.
I 23 Mike. i 1
1 24 MR. TRAVIS: The first slide. i 25 Our intention here was to give an overview of NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.
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40 1 the Westinghouse involvement in MOX fuel and then let Rick 2 talk about the technical details of our activities over 3 the last four years.in the MOX Fuel Program.
4 You know, Westinghouse's involvement, as Brian 5 said, goes back to the 1960s with MOX fuel. We fabricated 6 and irradiated lead test assemblies and partial core loads 7 of MOX fuel in European reactors, in the United States 8 also. We were involved in a commercial fuel fabrication 9 facility in the United States.
10 We do currently supply MOX fuel in Europe, and i
11 I'll talk about that in a minute, and then, of course, the 12 last four years we've been involved in this MOX fuel 13 study.
14 Okay. We fabricated the MOX fuel in the '60s i 15 and '70s and early '80s in a facility located just up the ,
16 street from where I work in Pittsburgh, Cheswick. -I spent 17 a. lot of time there personally in the '70s involved with 18 the design of MOX fuel and worxing with the fabricators. )
19 Brian mentioned our work in the Ginna n20 reactors, San Onofre, Saxton and Trino reactors. We had i 21 fuel burn-ups that at the time were equivalent to the UO2 !
22 burn-ups, and the performance was excellent in terms of ;
i 23 how they. performed in the reactor. ;
24 As Brian mentioned earlier, the fuel
() 25. fabrication processes were a little bit different than NEAL R. GROSS i COURT REPORTERS AND TRANSCR18ERS 1323 RHODE ISLAND AVE., N.W.
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41 !
?
1- they are today, the current technology. That does affect i-2 fcel performance. The burn-ups of these tests weren't
! 3 where we are today in Europe, but they were equivalent at [
4 the time to UO2 fuel. ,
f l 5 I know this is a subject for a-meeting next !
i 6 month, but I just wanted to mention a few things about our i e
I 7 Anderson, South Carolina, fuel f abric.ation facility. We :
8 envisioned it being built and operational in the 1980s.
j 9 It is something that went through the licensing I i
10 documentation. We applied for a license. We' continued 11 that licensing' effort, the dialogue with the NRC, up until 12 1982 when Westinghouse discontinued that operation. i i
13 But the funds were committed to the project.
l
!.;O 14 We had bought the land. We were on the move. That was 1
15 the state of the technology in the late '70s, and until 16 the government moratorium on-recycling and then the 17 project was dropped when it became apparent that we
[ 18 weren't going to proceed on a commercial basis.
1.
l- 19 Okay. Next. I
- 20 Okay. In the 1980s and up through today, l
21 we're working with European vendors to supply MOX fuel in !
l 22 Europe. We do the core design, the licensing work at our j l
23 offices in Pittsburgh. We build the hardware'other than ;
24 the fuel in the United States. We've used various l
!- . 25 European fabricators to build the fuel pellets and to l
\
! NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS j 1323 RHODE ISLAND AVE., N.W. l (202) 234-4433 WASHINGTON. D C. 20005-3701 (202) 234 4 33
~ -
9
- 42
! I complete the fuel assemblies, and then they're shipped to i
- 2 the reactors for irradiation.
4 t
! \._,
- - 3 We have experienced excellent performance from
- 1 4 this MOX fuel, and you know, we've supplied from the 3 5 United States uranium fuel, and essentially there have 6 been partial loads of MOX fuel in these reactors. !
7- Clearly, the performance is. excellent, and 8 we've not seen any problems, and I think another thing
, 9 that's important to understand from what we've learned is i d
10 that our methodology of designing reload cores is 11 consistent with what we've. observed from the core ;
l 12 performance.
7
) 13 Okay. Go ahead.
1 14 The last thing, and I'm just going to touch on
- 15 fabrication here for a second and we'll move on. During 16 this plutonium disposition study that we did for the DOE i 17 in the last four years, Westinghouse did go through with J
18 some help from other companies the design of a MOX fuel.
19 fabrication facility. We came through with a licensing i
s
, 20 strategy, cost estimate, and looked at the environmental 21_ impacts of that facility.
l 22 I think Mike mentioned that that's the subject !
}. 23 for a meeting in a month, and maybe we can dwell on that a !
24- little bit more. I 25 I will quickly run through some of the areas,
]
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l \
l 43 l' the background of Westinghouse, and what I'd like to do is 2 to turn the presentation over to Rick Ankney. Rick has a l
3 lot-of background in core design and in nuclear safety i i: l 4 analysis. He's been the primary focus at Westinghouse for I
-l 5 the reload core design, and he's going to give you some of I 6 the details of that design.
7 Rick.
l l
8 MR. ANKNEY: Thanks, Mike.
9 As Mike mentioned,.I'm an engineer in the i
10 Westinghouse Core Engineering Department. Over the last 11 few years, I've been involved in several different' studies 12 to look at how weapons grade Mox fuel could be utilized.in l 13 Westinghouse light water reactors.
l 14 Most recently the study that we did involved
.15 doing transition cycles from low enriched uranium core to 16 a core with a significant amount of weapons grade MOX.
l 17 The-plant that we chose as the reference = plant for our l
18 study was a typical Westinghouse four-loop plant, 193 fuel 1
19 assemblies, 3565 megawatt thermal core power level.
l.
f
-20 And one of the goals of the study was to i
\
21 evaluate what sort of hardware changes or changes to plant i 22 operation might be needed in order to accommodate this 23 fuel.
! 1 24 Next slide.
l
!- 25 In particular, what we wanted to do was to try q l ~
l l NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHoDE ISLAND AVE., N W.
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44 !
1 to characterize these core designs relative to the typical L 2 kinds of core physics parameters that we observe with low 3 enriched uranium fuel, and I'll be showing you some of ;
I 4 those parameters in a few minutes.
5 What I'm going to say is going to sound a lot !
6 like'what my colleague from Combustion Engineering said. !
7 In general, when we do safety analyses for these cores, we 8 pick a bounding set of parameters, and for the most part, 9 with some very few exceptions, the parameters that we have j i
10 calculated for these weapons grade MOX cores fall within i 11 the bounds that we typically assume.
12 Next slide.
13 The features of the study were several. We A
14 used an actual plant as a starting point for transitioning.
15 in the weapons grade MOX fuel. To the extent possible, we l
16 use the current plant configuration, same power level, 17 same control rod configuration, same control rod type, 18 same kinds of operating parameters, temperatures, power j 19 levels, et cetera.
20 And maybe most importantly, we try to use the l.
21 same. kind of fuel management that our plants typically i i i L 22 use. Most of our plants operate on 18-month fuel cycles l
23 with assembly discharge burn-ups in the range of about 45 l
24 to 50,000 megawatt days per metric ton, and we try to keep i
- (N g) 25 to that because that's the kind of cycle length and the NEAL R. GROSS !
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45 1 kind of operation that our plants are used to and have 7- 2 really designed their whole processes around.
'-] 3 Next slide.
4 This slide gives you an idea of the sorts of 5 plutonium throughputs that we were calculating for some of 6 these conceptual designs. There's four designs listed 7 here: the first transition cycle, second, third, and then 8 a full Mox equilibrium cycle design.
9 As you can see in the second column there, 10 this is the percentage of fuel assemblies in the core that 11 were weapons grade MOX. In this core there's 193 fuel 12 assemblies. So by the time we got to the third transition 13 cycle, we had almost half of the core was weapons grade
\) 14 MOX.
15 In our designs for a single assembly, we would 16 -- every rod would be -- if it's a Mox assembly, every rod 17 would be MoX, and then there would be some low enriched 18 uranium assemblies in the core as wel? for the partial MoX 19 cores.
20 In the equilibrium cycle, every assembly, 21 every pin was a Mox pin, and you can see the kinds of feed 22 region plutonium contents that we were assuming there. By 23 the third transition cycle we were loading about .69 24 metric tons, and in the full Mox core was 1.7 metric tons, 25 (G ) and since this is an 18-month cycle, on an annual NELAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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- . 46 4
1 throughput basis, in the last column you see the numbers
- 2 there. The full MOX core was using about 1.13 metric tons i
i 3 per year of MOX.
i j 4 Next slide.
J 5 We generated core models for.these designs and I
i i
6 calculated a lot of core physics parameters, power i
?
I 7 distributions, reactivity coefficients, et cetera. We 1 8 found that, in general, the power distributions were quite 9 well behaved and very comparable to low enriched uranium 10 core power distributions. I 11 Our peaking factors, F delta H and FQ limits l
- l 5
12 were met. F delta H is the hot rod integral power. FQ is i
4 13 the peak power density over the average power density of
,O j 14 the core, and in general, these conceptual' designs met i i
15 typical kinds of limits for these parameters.
1 t
j 16 You do notice that in MOX cores.there's a l J.
I 17 slight shift in the axial power distribution. We s
18 characterize that by a quantity we call axial offset, 19 which is how shifted the power is up or down from the core i t
i 20 mid-plane, and you find that with these cores, because of
- 21 more negative moderator coefficients and more positive 1,
4 22 boron coefficients, that you tend to see more power in the i
. 23 bottom of the core slightly, and this can be accommodated I i
24 within core designs.
~25 Next slide.
l 2 NEAL R. GROSS i COURT REPORTERS AND TRANSCRISERS j 1323 RHODE ISLAND AVE., N W. !
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47 1 When you have mixed uranium-MOX cores, like 7S 2 the partial MOX transition cycles that I discussed i'") )
3 earlier, what you do is zone the plutonium pins in the MOX 4 assemblies, and this slide shows a typical zoning pattern l
5 that we used in the study that we did last year.
6 The reason that you do this is to flatten tha I 7 power distribution in the MOX assemblies. Because the low 8 enriched uranium assemblies have a much higher thermal 1
9 flux, the pins on the outside of the MOX assemblies, if j 10 they're adjacent to a uranium assembly, would see high 11 powers, and so in order to flatten the power distribution, 12 a simple design solution to take of that is to simply zone I
_ 13 the plutonium content in the MOX pins -- in the MOX fuel i a
\d 14 assemblies.
15 And so on the outside of this assembly you see 16 there's generally low plutonium contents and on the inside 17 is higher, and this is an effective way of managing power 18 peaking factors and power distributions within the 19 reactor.
20 Next slide.
21 You do see some differences in the reactivity 1
22 coefficients in a weapons grade MOX or MOX core, and what i
23 drives this is the harder neutron spectrum that you have.
24 The reason the spectrum is so much harder in a MOX core is
( \
(j 25 that the thermal absorption cross-section of plutonium l NEAL R. GROSS l
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~
48
- 1- fuel is much higher. So the thermal' flux is lower, and so
- 2 the faster thermal. flux ratio is bigger, and this tends to l'
1
! 3' m.ake things-like boron worthless. The boron just simply t
- 4 has a harder time competing-for neutrons relative to the 1 5- fuel.
1 i
! 6 So the boron worth is maybe one of the i
7 significant aspects of the reactivity coefficient
! 8 differences that you see.
i
', 9 These differences are most apparent at
-10 beginning of life when the boron concentrations in the I i
11 core are high,.the critical boron concentrations. '
a i
j 12 Another key factor that you see, as my
{
r-13 colleague from_ Combustion Engineering mentioned, is the 2
{k 14 delay neutron fraction is lower, and.this can affect some 15 fast transients like rod ejection.
16 Next slide.
f j' 17 This is a pretty busy slide, and I'm not going i
} 18 to go over every number here, but I wanted to try to i i I 19 capture what some of the trends are in these designs. !
[
J. 20 The first two rows give the maximum peaking I'
21 factors that we calculated. The full LEU core is an !
- 22 actual core that operated with il uranium fuel,'and I've. !
{ 23 given the third transition cycle numbers.for the partial C !
j.
24 MOX. core, and then on the far right is the full MOX core.
j 25 You can see in the first two rows of numbers !
i NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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3 1
49
- 1. there that.the peaking factors are really quite comparable 2- to the low enriched uranium core, in-fact, a little less
-'(' v'O 3 in the case of the full weapons grade MOX.
4 One thing I should mention about these designs 5 was that we didn't use any integral burnable absorber in
'6 the MOX fuel, gadolinia or zirediboride or erbia. We used 7 discrete burnable absorbers, what we call wet annual 8 burnable absorber that fits into the guide thimbles of the 9 i fuel assembly and is not integral to the fuel, and that's I 10- an important aspect of the core design because it limits )
11 the amount of reactivity hold-down that you can have in 12 the fuel, and you have to make that up with soluble-boron, i
13 and so sometimes it leads to high boron concentrations, 1
14 which is really the next few rows there. l l
15 You can see that some of the boron E
16 concentrations are higher than in the full LEU case, and 17 this is because the boron worth is simply less.
l 18 In the full MOX core, what we did is we 19 assumed enriched boron in the coolant. Instead of the 20 usual 19.8 -- I think the footnote there has a typo in it, L -21 and it should be 19.8, not 18.9, is the natural content of l 22 B-10 --
instead of 19.8, we used 40 percent B-10 in the 23 soluble boron, and that got the boron concentrations down i
24 to levels that were quite comparable to what we're used to
() 25 in the LEU core designs.
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i 1
! 50 !
1 I mentioned earlier -- I skipped over the one t
!n 2 line on delayed neutron fraction. You can see that the i
l?
U 3 delayed neutron fraction decreases as you increase the 1
-.4 content of plutonium.
5 Moderate temperature coefficients tend to-be I
6 more negative, which is in general a benefit. Power 7 coefficients also tend to be more negative, which for most 1 i -8 accidents is a' benefit. ;
l l 9 Next slide. l l 10 Shutdown margin is one of the key i 11 considerations in MOX core physics, and it's a key l
l 12 consideration because the effectiveness of the control ;
i 1 l
l 13 rods is less. Just like the boron has a harder time i i
i 14 competing for neutrons relative to the fuel, so do control- '
1 15 rods. So the control rods are simply worth as much as )
l
, 16 they are~on a uranium' spectrum. 1 i
17 We were able to meet typical shutdown margin L 18 requirements for partial MOX cores with a significant-1 19 amount of'MOX, but the excess shutdown margin does reduce, i
20 and so for the full MOX core design, we would use probably 21 a more potent control rod material, something like l 22 enriched B 4 C in order to gain back the excess shutdown
! 23 margin that we typically see in low enriched uranium 24 cores.
25 Next slide.
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l 51 1 Now, this table shows some typical shutdown [
1 2 margin estimates that we came up with for these conceptual l
3 designs, and again, I'm not going to go over every number 4 here, but if you look at the very bottom line, that line i f
! 5 indicates how much excess shutdown margin there is above l 6 the requirement that's assumed in the' safety analysis. t 7 Typically for this plant, a 1.3 percent delta i a rho shutdown margin requirement would be used in !
i 9 transients like the steam line break, for example, and the f
10 bottom number there is the shutdown margin that was ,
p 11 calculated above that requirement. So it's excess l-12 shutdown margin.
f 13 And you can see that for the full LEU core we [
14 were about .68 percent delta rho above the requirement. !
l l 15 The partial MOX core, that's reduced to some extent 16 primarily because of the reduced rod worth.
l l
l 17 In the full MOX core, we .used enriched B.C !
18 control rods, and really more than gained back the 19 shutdown margin that you lose just because the control l
l 20 rods are worthless. That enriched B.C is worth more than f
21 sidmium and cadmium in the MOX spectrum 1
22 So that, again, that's another fairly simple 1 h 23 design solution, if the levels of MOX in the core are such .
l 24 that rod worth is reduced to the point where you need y 25 extra rod worth in order to meet the requirement. j NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N.W.
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i 52 1 Next slide.
(T 2 Now, generally when we do safety analyses for E]
3 these cores, we choose a bounding set of parameters and 4 t
4 perform the safety analyses with those assumptions. In l
5 general the MOX cores fall within those bounding )
6 assumptions. There are exceptions like the delayed i
7 neutron fraction, for example.
8 A typical delayed neutron fraction for a low !
l 9 enriched uranium core that's assumed in a safety analysis l 10 is larger than what we were calculating for some of these 11 conceptual designs. However, we did do some detailed 12 analyses of things like rod ejection and found that 7s 13 because of other compensating effects, like lower control j t )
l 14 rod worth, like higher Doppler feedback, for example, we 15 were able to show acceptable.results despite the lower 16 delayed neutron fraction.
17 Other accidents that we considered are listed 18 there, steam line break, loss of flow, boron dilution, and 19 so on. In general, in looking at these transients and 20 looking at the kinds of assumptions for key physics 21 parameters that we typically make as a starting point for 22 these analyses, we feel that we can show that we could get 23 acceptable results for these even in full MOX cores with 24 changes to things like the control rod material, if
+
rh \
( ,/ 25 necessary.
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l t
! S3 ;
i
! 1 Last slide.
s 2 'My concluding slide just sort of summarizes !
i 1
l 3 the conclusions of our study. We feel that with partial 4 MOX cores we can accommodate really a fairly significant 5 level of MOX without any significant changes in the l
l 6 reactor, reactor operation.
7 For full MOX cores, we would probably 8 recommend a change to a more powerful control rod 1
9 material, something. like enriched B.C. It would be useful 10 to use an-integral burnable absorber or use enriched boron i
l 11 in the coolant to keep boron concentrations to levels that 12' we typically see in reload cores today.
13 And maybe most importantly, these designs do 14 effectively-disposition the plutonium. The Plutonium 239 1
15 content is greatly reduced and the Plutonium 240 content i
16 is greatly increased, and so they do effectively l
l 17 accomplish the mission of this program, which is to i i
18 denature this material.
19 Mike. I 20 MR. TRAVIS: Okay. Rick went through his l 21 conclusions from the neutronics and the safety evaluation 22 of the study, and I wanted to add a couple more things.
l 23 One_islthat these hybrid control rods that Rick mentioned l
1
- 24 with the silver-tipped B.C are not new. Westinghouse
( 25 supplies these, and they are used in some plants, but we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS
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54 1
do use natural boron and not enriched ~ boron. What Rick 2 was proposing was an enriched boron in these control rods.
3 The other point I want to make is that, you 4 know, we have fabricated and irradiated MOX fuel for over 5 20 or 30 years, and it's performed well. It's something 6 that's done, and you're going to hear a little_ bit later I 7 today about some of the experiences in Europe, but this 8 can be done. It's feasible, and Westinghouse believes and 9 it's something that we supply.
10 So I conclude that a lot of the technical 11 issues with the MOX fuel have been solved, and'that's the 12 conclusion.
13 MR. WALLACE: Thank you.
14 Our next speaker is Chuck Paone from G.E.
15 Chuck is project manager of MOX Fuels Program for G.E. He l
16 has more than 29 years of experience with them and is 1 17 responsible for the coordination-of all their MOX program 18 activities related to existing plants, as well as related 19 to ABWR designs going into.the future.
20 Chuck.
i 21 MR. PAONE: Thank you, Mike.. '
22 In the 1960s and 1970s, prior to the change in-23 U.S. policy on recycling of reactor fuel, G.E. designed, 24 fabricated, and tested MOX lead use assemblies in the Big
( 25 Rock Point, Dresden I, and Quad City I BWRs and provided NEAL R. GROSS court REPORTERS AND TRANOCRIBERS ,
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, , ~,
. . - . ._ . . _ _ _ _ _ - ___.____ _m __. _ ._ _ . . _ . _
55 1 reload quantities of MOX fuel for the Gurigliano d
2 (phonetic) BWR in Italy. That experience and more recent
' (. 3 design studies provided G.E. with a significant baseline ;
i .
- 4-for designing weapons grade MOX cores and managing weapons !
?
5 grade MOX fuel cycles.
t
- 6 In 1966, an extensive study was undertaken for !
7 the DOE to evaluate the potential for improvements in the !
8 PU dispositioning rate and to optimize the fuel cycle so ;
?
l 9 as to make it more suitable for use in contemporary t
e 10 commercial reactor fuel cycles. This presentation is 11 going to briefly discuss the results of that. study.
12 The presentation is going to, as I said, N
13 briefly cover the BWR MOX fuel and core design. It won't i 14 get into extensive detail here; just tell you what the l
15 design looks like and what the results were. i 16 I'll then present our proposal for a MOX fuel 17 licensing plan both for LUAs and reloads; then briefly 18 describe the regulatory basis review we performed, design 19 modifications required, results of transient behavior, 20 qualification, and finally conclusion.
21 The design characteristics employed in this 22 study are, first ofEall, the use of a BWR/6 reactor plant.. l 23 This is our newest BWR design, and it employed the GE 11 !
24 (9x9) fuel platform in an equilibrium cycle core. I 25 We used an 18-month cycle length, 85 percent NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N W.
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56 1
capacity factor, which includes 45-day outage, a 33 2
percent reload batch fraction, which is about 264 bundles 3 per reload. The core we looked at was a 800 bundle core, l
4 and we developed two different MOX designs. One was a UO 2 5
look-alike and the other was the high throughput MOX 6 design.
7 I'd like to say that this set of parameters 8 for i reactor is fairly typical of the.way many BWRs are
-9 running this currently.
10 To make the licensing of the initial MOX cores 11 easier, the differences from conventional UO core designe 2
- 12 were minimized, and the initial approach for a MOX fuel 13 design was-to achieve a UO2 look-alike.
'14 In this design all of the MOX rods are placed 15 in the interior of the bundle and are surrounded by 16 peripheral UO2 rods. This MOX design behaves pretty much l 17 like a UO2 lattice.
18 UO2 look-alike design does not maximize the PU l
! 19 throughput and was proposed for initial MOX cores to l 20 establish the licensing basis for MOX fuel. These are 21 some of the details of this design.
22 So for the UO2 look-alike, again, as I
[ 23 mentioned, we used the GE 11 (9x9) mechanical platform, i
24 In all our designs we didn't use any gadolinia in the MOX i 25 fuel. There are UO2 rods next to the water gaps, and NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE N W.
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57 1
that's to shield the plutonium inside from the wide gaps 2 that soften the spectrum.
j 3
It consisted of 32 UO2 rods. Again, these.are .i 4 on the periphery of the bundle; 26 MOX rods; and 16 ;
i
- 5. gadolinia rods. The PU weight spread in the MOX fuel is f
6 2.3 to 5.4 percent. The plutonium loading per bundle is 7 2.45 kg total, 2.31 kg fissile. The loading per cycle is ;
- 8. .647 metric ton total, .61 fissile. Throughput per year 9 is .432 metric tons.
10 Next slide.
11- So to-increase the plutonium disposition rate, i
12 we then designed a high PU throughput MOX design. In this i
13 design, the fuel assembly is configured such that all but l
i
! 14 the gadolinia rods were MOX rods. Again, the details are l l
15 described on this page. Nine-by-nine designed the same, '
16 GE 11 mechanical platform; again, no gadolinia in the MOX 17 rods. In this case, all the rods, as I said, were MOX l
18 except for the 20 gad. rods.
19 The weight spread was 1.4 to 6.2 percent. The l
i 20 PU loading per bundle, 4.7 total, 4.43 kg fissile. Per l 21 cycle was 1.41 -- I'm sorry -- 1.241 metric ton total and 22 1.17 metric ton fissile. The plutonium throughput per l 23 year increased. It was almost doubled to .827~ metric tons.
i-l 24 total.
25 What we had in mind was to start the reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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58 1
with the UO look-alike and than transition to the high PU 2
gr'$ 2 throughput after the UO, look-alike was in the reactor for
%j 3Ja few cycles. l
'l 4
This slide shows the results of the summary of 5 the equilibrium MOX core design. The hot excess 6 reactivity cold shutdown margin and thermal limit design 7 targets are satisfied. We've met all of the criteria !
8 essentially. !
i l
9 We have a negative void, power, and Doppler 10 reactivity coefficients as is required. We have a 11 negative moderator temperature coefficient at operational 12 temperatures. It's slightly positive in the start-up l
-s 13 range, but that's allowed per the criteria.
I I
' 14 The stability criteria were met. The high MOX l l
15 core design has a smaller flow window. I don't know how 16 many of you are familiar with BWRs, but there is an 17 exclusion zone at the lower flow ranges, and this 18 exclusion zone gets a little larger. I'll discuss that a 19 little bit more later.
20 This provides some additional information on 21 the MOX core design. I'm not going to get into a lot of 22 these numbers. I'd just like to note, again, the 23 plutonium throughput per year for the UO2 look-alike bundle 24 is .432. For the high MOX core, as I said, it's almost
/
(,)N 25 doubled.
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59 1
And I'd like to note that we've done or been 2
involved in some work in developing even higher throughput y
3 designs which would result in a yearly MOX throughput of 4 1.6 metric tons per year. So, again, essentially almost 1
5 doubling the high MOX design.
6 However, that design requires burnable poisons 7 in the MOX fuel.
8 I'd like now to talk about BWR MOX fuel 9 licensing. This slide presents a flow chart for the 10 proposed MOX LUA licensing program. In this program, we 11 propose to use 10 CFR 50.59.
12 We would start with a description of the LUAs.
13 That includes the mechanical, number of gad. rods, O
k/ 14 standard GE with a few segmented rods if at that time we 15 decide that we'd like to bring some of the segmented rods 16 back for PIE work. MOX rods would be built to GE specs.
I 17 and QA requirements. '
18 After that description, we'd :3 tart a methods 19 validation program against existing data to insure that 20 the methods provided accurate results; go through a design 21 review of the methods validation. I 22 In parallel, we'd have a technical update 23 meeting with the NRC to describe the process we were 24 proposing to use and basically go through our standard
() 25 50.59 performance evaluation, resulting in loading of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N.W.
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S P
60 1 LUAs. l i
10 CFR 50.59 allows changes described in the
()
2 i
1
= CAR if there are no unreviewed safety questions and no I i
4 tech. spec. changes required. The LUA is the only tech. ;
5 spec. change we anticipate, would be a minor change to 6 acknowledge _MOX fuel. In some cases, tech. specs. specify 7 UO2 ' f uel .
8 The unreviewed safety questions are defined 9 as: is the probability of an event increase? Is the 10 possibility of a different-type of event created? Are the 11 margins to tech, spec. bases reduced?
12 We would review these criteria in detail, do 13 evaluations, and at this time we believe that the MOX.LUAs ,
iO 14 would meet these criteria easily. There essentially would i l
15 be no unreviewed safety question.
16' We would provide a 10 CFR 50.59 safety 17 evaluation report. This would define the number of LUAs 18 to be inserted, list of new features of the LUAs, and 19 'xamine the impact of each new feature on the FSAR event, i
20 such as the LOCA, ATWS, control rod drop action, et 21 cetera.
22 10 CFR-50.46 states that light water reactors 23 fueled with uranium oxide pellets with cylindrical 24 zircaloy or zirlo cladding must be provided with an ECCS.
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4
, 61 l 1 the fuel that we're talking about here is obviously not n 2 uranium oxide. So, again, it must acknowledge MOX fuel.
t 3 The deviation report must show that the MOX l
4 3 4 fuel will not significantly impact these items listed, the I 5 peak cladding temperature, maximum cladding oxidation, 4 ;
6 maximum hydrogen generation, coolable geometry, long-term 4 7 cooling. i i
j_
8 Again, these criteria would be analyzed in 4
! 9 detail, and again, we don't anticipate any problem, any i
- 10 concerns resulting from that evaluation, especially for l 11 LUAs. j 4 .
1 12 Both reports would be submitted to the NRC for !
13 review. 1 I
4 14 Next slide.
j 15 This slide shows the MOX LUA licensing i
l 16 schedule. It's 18 months from initiation to completion.
j 17 We realize it's somewhat optimized, but believe it's i 18 doable, .and it just goes through the items I just i
19 described.
1 i
j 20 Of course, we'd request NRC approval for this 4
11 21 approach prior to even starting on it, and that's the 22 purpose of the initial meeting with the NRC.
23 This is a flow chart for BWR MOX fuel j 24 licensing for reload quantities. We would use our 4
l 25 Amendment 22 evaluation. I'll describe that in a little f
i
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d 62
[ 1 bit of detail just in the next slide. At the'same time,
- 2 we would initiate our BWR MOX bundle design, submit issues 3 to the NRC, obtain NRC approval, and then do our standard
- 4. core analysis process, resulting in the plant specific 5 supplemental reload. licensing report.
3 6 So for MOX. fuel reload core licensing, we 7 propose to use Amendment 22 to our GESTAR document.
8 Amendment 22 contains a-set of fuel licensing acceptance 9 criteria that have been established for evaluating new 10 fuel designs. Fuel designs which meet these criteria 11 constitutes U.S. NRC acceptance and approval for use in 12 BWRs.
13 Amendment 22 licensing acceptance criteria 14 state that NRC approved analytical models and analyses 15 procedures will be applies. New design features will be 16 included in lead use assemblies, which is what I just i
17 described. The generic post-irradiation examination )
18 program approved by the NRC will be maintained. That 19 could.be in two parts, one in the LUA program and second
'20 with the actual reload' quantities. We'd take a look at 21' some of the bundles after each outage or after each cycle 22 to determine if there's_anything unexpected.
23 -If.any of the fuel licensing criteria are not
~24 met for a new fuel design, that aspect will be submitted
() 25 separately to the NRC for review. So basically as part of NEAL R. GROSS :
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~ - _ _ _ _ . _ _ _ _ . _ _ . _ . . _ - . . _ . . _ _ _ _ _ _ _ _ _ . - _ _ . _ _
63 1
the Amendment 22 evaluation process, issues'related to MOX 2
fuel that do not meet _the criteria or are questionable O '(
3 will be identified and submitted for review. :
4 'This slide shows the anticipated MOX reload i 5 licensing schedule. It's essentially four and a half 6 years from start to finish. ;
7 The core analysis section may be a little 8 long, but we included some extra time just_because first 9 time through on these things, as all of you know, it takes 1 10 a lot longer, but four and a half years, I think, is a 11 fairly' reasonable goal.
12 Next slide.
13 I'd just like to shift to a real brief O 14 description of some of the things we've done on various 15 other areas. One of them is regulatory basis, and this 16 chart shows the result of the regulatory basis review. .
17 First of all, I'd like to say_that we should 18 utilize past MOX experience. There's a lot of information I
19 available from past work, both domestically and in Europe, 20 and you know, licensing related, design related, 21
~
experimental work, and we should utilize that in this 22 program. We really don't have to reinvent the wheel here. I 23 We reviewed 10 CFR 50 in a general' design 24 criteria and found nothing that prevents the use of MOX l
( 25 fuel as long as all the criteria contained in those NEAL R. GROSS
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64 -
l 1 documents are met. We reviewed in detail the technical 4
l
\
2 specifications and FSARs. We found some minor I t
3 i modifications needed to acknowledge MOX, as I previously j 4 mentioned. In some cases UO is specified, and we'd have 2
l I
5 to change that. .
6 There are no unresolved safety questions l 7 expected. We really don't feel there's any real technical l
i B
issues here, and we feel we can obtain NRC approval under i 9 10 CFR 50.92. l S
10 We reviewed the BWR plants for required-design j l
11 modifications, and our review indicated that there are no- l
?.
12 major' modifications needed. All control functions i f 13 addressed in fuel design -- are addressed in the fuel O 14 design, and by Amendment 22 provisions. So basically when 1
l 15 l the fuel is designed, we have these set of criteria that :
16 have to be met, and in doing the design, the engineers ,
17 meet those criteria. So, you know, there is essentially 18 no real requirement for any major modifications. -l 19 Some minor modifications will be required, l t
20 including stability monitor set points and new fuel i 21 storage security. As I mentioned, the security monitor j
! i
' 22 set points are to accommodate a more narrow, low control !
l l 23 window. .The exclusion region at the bottom end of the ,
s i
i 24 power flow map will have to be increased a little to '
o I
() 25 accommodate MOX fuel.
i '
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1 1
65 l l
1 An important point to note is that the costs
(
f~s I
2 associated with these changes is really not significant.
i/
3bA 1:t of it is electronics. There are no real major i
4 changes in any of the major equipment in the BWRs. !
l 5 Trancient behavior. The MOX impacts the 6 operating limit CPR, and that will be slightly affected.
7 The operating limit will increase to some extent. The 8 more negative void coefficient results in a higher delta 9 CPR, which will, in turn, cause the operating limit CPR l 10 being a little higher than in the UO2 core.
11 This will not create an operational difficulty 12 since operating rod patterns will be developed to
,_ 13 accommodate the higher operating limit. We don't
/ i
\~ / 14 anticipate this to be significant.
15 As I mentioned, the instability, the exclusion 16 areas impacted, but the operational impact is expected to l 17 be minimal. It will affect the way the plants are started 18 up. Instead of being able to go to a relatively high 19 power at a given flow, the rod line used will have to be 20 lower.
21 Another area we examined is MOX fuel j 22 qualification. We developed plans and recommendations for 23 fuel design methodology, that is, to validate codes and 24 methods against existing data. Plutonium metal to oxide
( ) 2e conversion to assure the output of the plutonium metal to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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66 .l 1 .!
oxide process is compatible with the MOX fuel process, and i i
2 in the MOX fabrication process itself, to assure that the ;
I 3 pellet rod and bundle fabrication meet specifications. l 4-i Finally, in conclusion, G.E. BWR MOX fuel !
5 designs allow immediate implementation, no major impact on 6 plant operations, and PU throughput up to .83 metric tons !
\
7- per year, and potentially as high as 1.6 metric tons per t i
j 8 year.
i l
9 The licensing is compatible with the 10 established process, and there are no significant plant l 11 modifications required, and we also have a MOX fabrication I 12 qualification plan and experience in place. ,
i 1
l 13 So, in summary, we feel G.E. BWRs offer solid N
14 opportunity for safe, effective, and early PU disposition.
i 15 MR. WALLACE: Thanks, Chuck.
16 Ceuld we bring up the lights now because we'll i 17 go into the question period? And I think, Vanice, what 18 you preferred is for people with questions to come around 19 to the microphone, okay, so the court reporter can hear.
20 We'll look to the NRC staff first of all since 21 this is primarily an open technical meeting to engage in l
22 discussion between the staff and the industry. So I would 23 look first to the staff for questions you might have of l 24 any of our presenters here.
( ) ;25 Come right on up. Take the lights the other NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS i 1323 RHODE JSLAND AVE., N W. I I
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- i. ,. __ - ,
.- . . - . - - . - - . - . - - , . . - . . . - - - . . - - - - . ~ _ - - - ~ . - . . .
l 67 l
- 1. way. l i
g 2 (Laughter.) i
\ 3
~3 MR. WILLIS: I'm Charlie Willis, NRC, and you l 4' can leave the lights down if you'd like. I'm not going to 5 win any beauty contests anyway.
]
1 6 (Laughter.)
i 7 MR. WILLIS: Questions concern the fission 8 product inventory and release in a hypothetical accident.
9 Do we have fission product release data for the mixed 10 oxide fuels, and have we done the calculations to deal !
'11 with the higher inventory of Iodine 131 that would be
- 12 expected in a mixed oxide fuel?
l 13 MR. WALLACE: I'd like.to ask each of the l
! 14 vendor company ~ presenters to address that, and we'll start 15 right in the order in which we went. So, Stan, if you'd~
16 go first.
I-
'17 MR. RITTERBUSCH: Stan Ritterbusch, ABB.
I 18 I would say we have not done the specific 19 analyses that were requested. I expect there'are many l
[ 20 people within NRR who can answer that question in much i
21 more detail than I can.
22 I do note, however, that the new source term i 1
23 technology that we would use, that is, that documented in
! 24 NUREG-1465, has a substantial amount of conservatism in it j()
k 25 with respect to the experimental data and the analyses I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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- . - . . - - - . - - - - - - - . - - - .- ~ . .~. - .- .- - . _ -
I 68 1 upon which that report was based.
l 2
1 So I would hope that the answer to that ;
3 question is that the current deterministic source term 4 release methods bracket that, but of course, that would be !
5 subject to somebody confirming the specific details.
6 MR. TRAVIS: We did take a look at the.' source.
7 term from all the isotopes from MOX fuel and compared them 8 with UO , and we did find that, as you mentioned, Charlie, 2
9 the isotopes for iodine were higher, but there were other 10 isotopes that were lower, and iodine is the most important 11 one obviously, but we did take a look at the source term 12 and found, as Dan said, that it appeared to be within 13 limits.
l 14 But I think that's a subject for a much 1
15 detailed, further study here.
! 16 MR. PAONE: Yes. We, too, took, a look at the l 17 source terms and found what everybody else has found.
t 18 Basically there isn't much difference, however, and we 19 don't feel there'll be much of an effect.
i 20 We had a table that was included in our report i 21 to DOE. 'So that's available for review.
22 MR. WALLACE: Okay. Additional questions? !
23 MR. ARCHIBAL: I name is Ralph Archibal with 24 NRR. !
)
] !
- 25 I had a question on the question of, I guess, '
i NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N W. l
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_._ _ - . __ _ m . _ _ . . . . _ . - - . _ . _ _ _ . _ _ ~ _ - . _ _ . . . - _ _ . ._
t 69 i
I the general approach and strategy. I was under the t
~2 impression that these would as a minimum be license
' [~)
\_/- >
i 3 amendments, according to the expectation of MOX fuel, and c i
4 the General Electric presentation kind of raised some !
i l 5 thoughts in my mind that there might be some idea that the -
6 industry was proceeding with the thought of doing it under I 7 Provision 50.59, and I don't really -- I didn't want to i
8 endorse that in this meeting, but I have the impression l 9 that any action along these lines would be at least 10 initially licensing action required, amendment approval, i 11' et' cetera, and we'd get into a full review of any initial '
i 12 use of MOX, and I wanted to raise that point.
i !
13 Unreviewed safety questions are currently i
14 being examined right now. We're reexamining the basis for i
- i 15 ,50.59 in those evaluations. So that part of that j 16 presentation kind of surprised me a little bit, I guess.
I 17 MR. WALLACE: Okay. I might ask again each of ;
l 18- the vendor companies. I recognize that could as well be a i
I 19 question directed to the licensees of the light water ;
L i 20 reactors, but with the focus for the study that has been l l 21 done by each of these firms, we ought to ask you guys to 22 comment 23 S' tan. i
. \
24 MR. RITTERBUSCH: I would agree surely there i
- ) 25 would be a change to the technical specifications as each NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS l- 1323 RHODE ISLAND AVE., N W. I (202) 234 4 433. WASHINGTON, D.C. 20005-3701 (202) 234 4433 l
1
- . . - .... .- . . . . . . - . . _ _ . . _ _ _ _ . - . - - _ - - . - ~ . -
70 1 of us have said. We all observed what is happening with 1
i 2 the whole 50.59 process, and we realize there would be a I
(
3 significant number of questions by the-NRC staff, and we i
4 would have to just follow the appropriate process.
5 MR. TRAVIS: Westinghouse took the tact that 6
we would' license MOX fuel as part of the manufacturing l
7 license, the end use of the MOX fuel, and the unresolved i
B safety questions would be addressed at a time when you got 9 a license for the fuel fabrication facility, and these i
10 issues would be addressed before. you committed to l 11 fabricating the fuel.
l 12 MR. PAONE: I would just like to emphasize the 1
l 13 proposal for use of 50.59 was just for LUAs, to-hopefully {
14 implement LUAs on a fairly quick basis. It really had 15 nothing to do with licensing for reload quantities, and in 16 reality the proposal would be partially 50.59 but include 17 the NRC right from the start and to a much greater extent 18 than a normal 50.59 evaluation. l 19 MR. ARCHIBAL: Ralph Archibal, and I had a I 20 follow-on questien to Charlie Willis on the -- I think we i 21 raised this in a previous discussion about the testing or-
'22 any testing that was planned in the source term area for 23 the release, the chemical aspects of the source term, j l
24 which are all based now on OFF, et cetera, type 1
- 2 5' experiments without this type of fuel.
j I
i NEAL R. GROSS :
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71 1 Is there an experimental program planned to 2 address the effects of chemical and physical versus !
3 isotopic on fuel performance during accidents?
4 MR. WALLACE: I'd ask any of the speakers if 5 you care to address that.
6 MR. PAONE: We have none, no plans at this
- 7 time.
8 MR. WALLACE: Okay. Next question?
9 MR. EBERT: I'm David Ebert from Research, NRC 10 Research.
11 I have a question about for rod drop accidents l 12 or rod ejection accidents. You keep talking about that 13 the reactivity worth of the rods'is smaller, but that O
i i g r.i 14 depends on what units you give to it. So I suspect that 15 the rod ejection or rod drop would still produce a-prompt, j 16- critical reactivity. l l l 17 My question is have you looked'about the 18 energy deposition for rod drops or rod ejections. Is it )
J 19- much different than the present day uranium type course?
l' i 20 MR. WALLACE: Stan, we'll start with you. l I 2 l 21 MR. RITTERBUSCH: We did not repeat the i
22 analyses for this particular study. However, we-have done !
i 23 some very specific analyses in prior years for recycled i 24 MOX cores. We did look at those events, and you are i
$ 25 stating it correctly that the energy deposition is not l NEAL R. GROSS
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m r ---a y 4 4 -m,a v -- - - -
_ . _ _ .-.-.__.____.____..________m__.. _ _ _ _ . _ _
4 i 72 3 a
j -1 much difference. i~
i 2 There is a prompt, critical power peak for
(
, 3 certain of those events. However, the fuel temperature t
] .
4 coefficient, you know, cuts it off quickly, just as with a f s
5- mixed oxide core.
. I 6 MR. WALLACE: Rick. .
y 7 MR. ANKNEY: This is Rick Ankney, f 8 Westinghouse. 'i 9 We did do explicit calculations of the rod l
r
- 10 ejection event, and I think it's included in our report. l a
11 We did see there is an effect of the lower delayed neutron '
4 .
i 12 fraction. You do see in some of the cases a higher opike, 13 but the total integral power is comparable to or less than j t ,
-14 what we typically see in LEU course, and that's because of ;
1 j 15 other compensating effects like lower control rod worth, .
a j .16 higher feedback effects, and so on. 'i 4
e 17 So overall when we looked at the transient, it
- l. 18 was really fairly benign. i 4
l 19 MR. PAONE: For BWRs, we didn't do a specific 4
20 control rod drop accident, but in BWRs because of our
, 21 control systems, the incremental rod worths are really 4
22 low, and although the event would be prompt, critical, the ,
23 extent of the event, the damage is really pretty low.
(
i j 24 MR. WALLACE: Okay. Additicaal questions? ,
25 Members of the staff?
.i
- NEAL R. GROSS i COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W. *
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73
- 1. .Okay. For members of the public?
2 MR. LYMAN: I'm Ed Lyman, Nuclear Control Hi.
,'(
l 3 :nstitute.
4 Just a follow-up on the rod ejection. All the 5
participants stated that it was essential that or a part 6 of their analysis that they maintain current models of I 7 core management, and in particular, the burn-ups of the 8 MOX fuel assemblies would-be comparable to that already 9 used for uranium fuel.
10 But in view of the recent CABRI tests in which 11 they found a violent rupture of MOX fuel that was burned L
12' up to around 55,000, I believe, taking that into account, 13 plus the maximum burn-ups that you assume-for MOX fuel,
)
14 how would that affect your analysis of, for instance, the 15 control rod ejection, in which case you might have'a !
16 change in the number of fuel rods which violently rupture 17 and might clog control channels and block the possibility 1 18 of -- how would that change.the analysis you've already 19 done for control rod ejections?
20 MR. WALLACE: Okay. I guess I'd ask each of i
! 21 our panelists to address that again, if you would please. !
i . l
[ 22 Stan.
l 23 MR. RITTERBUSCH: The issue with respect to 4
24- the tests that were referenced has to do with the 25 potential of lowering the acceptance criterion for the
]
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74 ;
1 particular accident. Of course, that has to be addressed z3 2 in the particular analysis.
h 3 There has been a substantial program in the 4 industry, and you know, the NRC has evaluated that. My 5 colleagues may know the results of that or a current 6 status. Possibly they can answer more cogently.
7 MR. ANKNEY: I guess I'd just like to point 8 out that in our core designs we were pushing the MOX fuel 9 to burn-ups of about 45,000. Okay? So we weren't getting 10 up to a level of the 55,000.
1 11 Obviously you have to be able to show that the '
l 12 fuel will operate normally at high burn-ups, and if we l
,_ 13 were to push the fuel that hard, we would have to put into 14 place the analyses and programs to show that the fuel 15 integrity would be maintained.
16 I think at the levels in these conceptual 17 designs that we put forth in the study, we would expect i
! 18 the fuel to operate fine.
19 MR. PAONE: Our fuel is taken out to 45,000 20 also. So if you want to have some correlation, it will be l
21 well below the exposure at which that failure occurred.
22 I'd just like to say that we ought to find out 1
23 a little bit more about the failure before we start to 24 draw some conclusions here. Fuel out that far, normal UO 2 (g,) 25 fuel has got a lot of MOX in it also, and there may be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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! -J J l 75 I 4
1 other considerations relative to cladding or other things i
j I
j 2 that may have affected the results of that test i j 'l
. 3 MR. WALLACE: .Okay.
4 MR. MEYER: I'm Ralph Meyer from the NRC 'I 1
! .5 Research staff. i 1
i- I 6 I can tell you a little about'that test.
i The 7
recent test in CABRI that did experience a cladding
, ~
l 4
8 failure was, in my opinion, a very successful test,~and 9 the failure.was exactly what we expected. It was -- as
- 10 you have indicated, there have been some apparent changes
- -11 needed in the criteria.we're using for analyzing this
, 12 accident for high burn-up fuel rod cladding, and we have
[ .
13 been assessing this situation for-a couple of years now.
l 14 and see that in most cases the threshold for cladding
{. 15 failure is occurring around 100 calories per gram for PWR 16 cladding with a fairly heavy oxide coating on it-.
I E 17' This fuel rod with the mixed oxide fit ~ the !
1 1
18 pattern exactly, and I don't think there's anything ~
l 19 alarming about that' test, and I was very happy.to see the f 4
20 result because it fit the understanding that's emerging 21 quite well.
i 1
1 22 MR. WALLACE: Thank you. I appreciate that V
23 additional information.and clarification.
24 Other quests.ons from anybody in the audience? l
,. q -
3 Q 25 Okay. I'll declare the end of this session.
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I' l 76 1 It's by my watch 10:40. If we can result at 11 o' clock, 1
.{-(
2 please, then we'll go into the session on European 3 experience.
] 4 Thank you.
5 (Whereupon, the foregoing matter went off the 6
- record at 10:48 a.m. and went back on the ,
3 j I 7 record at 11:11 a.m.) I 1
8 MR. WALLACE: Okay. .If you don't mind, we'll T
l 9 get started on the next portion of the meeting, if I can d
10 get a bit of a call to order.
l j_ 11 A logistics item that I'd identify first --
1
- . 12 thank you -- a logistics item. Please make sure that i
- 13 you've signed up on one of the sign-up sheets on the table i g j 14 at the entrance to the auditorium. By so doing, we will i
- l. 15 be sure that you get a copy of all of the presentation
, 16 materials that are presented here since we didn't have
- 17. copies of all of them available prior to the meeting. So 18 your name on the list will assure that you get a copy of !
19 the materials. !
1 20 As with the last session, we'll follow the '
i 21- same format, our two speakers first, and then we'll 22 entertain questions related to the presentations that they 23 will be making. Following that, we then.will open it up 24 generally for questions of the entire panel.
l 25 The first speaker is Jean-Luc Provost from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W. !
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I 77 1 Electricite de France. Jean-Luc is part of the' Fuel 2 Operations and Strategy Group. He is in charge of the 3 engineering activity in the Nuclear Fuel Branch. He has 4
19 years' experience in the French nuclear industry, 5 involved in safety studies, PWR operations, and fuel 6 procurement.
7 Jean-Luc.
8 MR. PROVOST: Thank you. Thank you, Mr.
9 Wallace.
10 Ladies and gentlemen, the purpose of my 11 presentation is the EdF. operating experience concerning I
12 MOX fuel in PWR. EdF is a French utility supplier, and it L13 operates about 55 nuclear power plants today.
14 The content of my presentation is with six 15- items: historical outline and our strategy. The second 16 is adaptation of PWR plant for recycling. The third is 17 fuel design and core management. The fourth is the EdF
- 18. experience. The outlook for EdF and some main 19 conclusions.
20 The historical outline. In France, closing 21 the fuel cycle is a constant policy, and French nuclear 22 fuel has been designed to be recycled.
23 In 1985, the decision is taken to recycle 24 plutonium in the PWR on a large scale, and in the
() 25 possibilities of recycling in PWR has been demonstrated.
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78 1 The first experiment of MOX, like has been said, is in the 2
Belgian reactor, BF-3, in 1963, and from 1974, we dispose O 3 Of an experience of recycling in the French reac' tor Chooz 4 A.
5 And the third important item is that the 900 6 megawatt plant has been designed to receive MOX at the 7 origin, and to close off the fuel cycle, there was another 8 decision in 1992 to recycle, reprocess uranium in our 9 plants.
10 The current strategy, two aspects. The fluid 11 quality strategy, it consists that the amount of separate 12 plutonium will.not exceed what can be manufactured and 13 consumed in reactor, equilibrium between produce and O( / 14 consume plutonium. j i
15 And the second goal is that the MOX-uranium
{
1 16 parity regarding utilization. MOX must provide the same 17 nuclear power plant performance that uranium in terms of l
18 energy and in term of reactor operation. It's an 19 objective.
20 The second part, adaptation of PWR plant for 21 recycling. Four aspects are examined: reactor 22 adaptation, MOX transport and handling, fresh fuel 23 checking, and spent MOX transport.
24 The adaptation of the plant. We must take
() 25 into account the specificity of the neutron spectrum with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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3 79 ,
- 1 1 plutonium. Like explained this morning, we'must take into I n !
2 account lower efficiency of the absorber, the boron, and i 3 the control rods.
? 1 4 The impact on the reactor is three main j ,
(
5 impacts. The, reactivity shutdown margin decreases. Its ;
l 1 6 impact acts on transients studies. The safety injection 1
7 efficiency decreases. In particular, the main impact is 8 on the steam pipe break accident in our safety analysis 9 report.
10 And the third impact is the boron make-up
- 11 system capacity is short, in particular, for shutdown I >
12 situation, to pass from power situation to shutdown 13 situation.
. \
j 14 And the adaptation. We increase the number of i
i j 15 our CCE. We use four extract on rod plaster (phonetic), l
.16 57 instead of 53, and we increase boron concentration in I j'
'17 the refueling water storage' tank, the value past 2,500 t .j 18 ppm, and in the boron make-up tank to 7,500 ppm.
i 19 Here you can see the RCCA position with old 20 location of the rod, as the values indicate, and-when we 21 change the core management from the oldest core management i
22 to the actual core management of uranium, we transferred 23 four rods at the outside of the core. It's ex, outside, i 24 and when we implement the MOX core management, we increase 3
25 the number. We take the four locations in the center, the
?
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80 !
i i black'four locations in the center in the new location of '
2 rods, four new locations.
3 Next. I 4 MOX transport and handling. For the fresh 5 fuel transport we need to design a new container to take 6 into account mechanical and radio protection device for 7 the specificity of the MOX fuel, and due to this device, 8
we have a heavier container which needs to reinforce the 9 handling crane in the fuel building, i
10 The next aspect is receipt of the fuel in the 11 fuel building. We have a constraint. It's to receive two 12 weeks-before shutdown the fuel, to limit the time with
- 13 fresh fuel in this building, and the reason is that.we l l 1
lw 14 have access restriction of this building during this' 15 period, and-today we' don't have authorization to know 16 device duration due to radio protection and strategic l- 17' reasons.
18 Another aspect is handling. The procedure has j' 19 been modified. No movement of assemblies at more than 70 t
i 20 centimeters of the floor due to the risk of assembly drop,
=
r 21 and operator number limited in the fuel building during I' 22 handling. Today it's eight persons maximum.
l l 23 The experience of ten years of operation gives i
l 24 no particular program for transport, receipt, and 25 handling.
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i
!~ . _ .
81 1 The fresh fuel checking. Specific controls r .
2 were added before to the container, in particular, control
, \
3 of temperature, of gamma, and reutron measurement, and we 4 implement special cells to examine the fresh fuel using i
'S radiation protection, particular system to reduce the q 6 doses, and examinations of the fresh fuel were limited to 7 the strictly necessary ones.
8 The procedure concerning information access, 9 physical protection were adapted to satisfy EURATOM and 10 the French safety authorities concerning plutonium up to 11 the close of the vessel. l 12 The spent MOX fuel transport. The specificity
. 13 of the MOX modified cooling, as a minimum cooling period l
[\' '!
14 in pool for'MOX is 2.5 years. It's one year for uranium, t 15 There's a potential problem of spent fuel size due to this l
16 increase.of the period of cooling and its problem of
'17 storage center or pool reacting. To date it's in 18 Studienne (phonetic), France, for the future.
19 The transport aspect. Fuels are transport -- l 20 MOX fuels are transported in standard casket for MOX fuel. I 21 In the center, the cask is surrounded by eight uranium 22 spent fuel, and we dispose of transport feedback. To date l
l 4
23 ten shipments have been completed at the end of 1996. The l 24 first shipment has been realized in the beginning of 1996.
() 25 The fuel design in the core management. Today NEAL R. GROSS
, COURT REPORTERS AND TRANSCRIBERS l 1323 RHODE ISLAND AVE., N.W.
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82 1 the fuel design used in our plant is a stand-out design.
l fg 2 This is a FRAGEMA advanced type AFA-2G. It's the same as b 3 for uranium fuel. The plutonium oxide is mixed with
(
4 depleted uranium, with an enrichment of .22 percent, to 5 concentrates of a maximum of plutonium in the minimum of 6 assemblies with choice depleted uranium.
7 We need to take into account in the fuel l
j 8 zoning. That is explained this morning. It's needed to l
9 balance' rho forward (phonetic) distribution at the
.10 interface between the uranium and plutonium assembly, its l
l 11 neutronic effect, and we have three zoning very close to l
12 the Westinghouse presentation of this morning. The first 13 one is the three rod in the corner. The second one is the 14 peripheral of the fuel, and the third zoning is with 15 maximum assay of plutonium in the center of the fuel.
16 Another aspect is that MOX fuel can be 17 reprocessed in a large factory, and we impose solubility 18 criteria on the pellets, on the spent pellets. The MOX 19 reprocessing feasibility has been demonstrated in 1992 20 with five ton of heavy metal reprocessed.
- 21 Now with the isotopic composition, today we
-22 use plutonium with about 70 percent of fissile plutonium.
l 23 The isotopic composition depends on the origin of 24 plutonium, and we can consider two types of plutonium in i(
J
) 25 the study. the first type is Plutonium 1 in the slide j NEAL R. GROSS l COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N W.
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- 'I 4
83 1 corresponding to spent fuel, uranium, with enrichment of 2 3.25, is former core management in France in the 900
- 0 3 megawatt plant, with burn-up of 33 gigawatt day per ton, i
4 five years of cooling of the spent fuel, and two years of 4
5 storage of plutonium after reprocessing.
6 And the second type of spent fuel that we must 7 consider today is spent fuel -- is plutonium with origin 8 spent fuel - .recent spent fuel with new core management 9 is management with 3.7 enrichment and with burn-up of 45 10 gigawatt day per ton, with nine years of cooling and three !
11 years of storage.
12 And if we want to obtain the equivalence in 13 terms of energy with these two types of isotopic 14 composition, for the first we need an assay of 5.3 percent 15 of plutonium and for the second we need to increase the 16 content of plutonium to 7.1, and we don'.t have the 17 authorization to date to the second isotopic composition, 18 but the authorization is today in discussion with our 19 safety authorities.
20 The next slide is an example of different 21 isotopic compositions that we used in the past in the
~22 first table, and you can see on this slide that the 23 isotopic composition is degraded during the time with the 24 origin of the plutonium. At the beginning, contain j
() 25 plutonium from reprocess of spent fuel at low burn-up from NEAL R. GROSS :
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. _.- ..---- - - ~- - - - - - - ~ ~ - - - - - - '
-"-^"T I
L 84 j i
1 l 1 PRW and also contain plutonium from GCR, French GCR, and '
l 2 it's a good quality of plutonium.
\
3 And' after in 1994 is bad plutonium with l
4 fissile content of 67. percent, and we don't have with this 5 type-of plutonium the equivalence. That's on the right 6 side. We don't the equivalence with the enrichment of 7 2.35 percent, is the reason that we try to justify to the 8 safety authorities that we can use the newly made of 7.1 9 percent.
10 Next.
11 So core management. Today in France the 12 recycling rate, the quantity, maximum quantity of MOX 13 assembly in the core is limited to 30 percent. It is due-14 to the effect of plutonium on the efficiency of our 15 system, and we have implemented in the past two types of 16 core management.
17 The first is third core management with reload 18 of -- each reload. It's 16 MOX equivalence 3.25 with 32 19 uranium fuel enriched at 3.25 percent, and the burn-up of i 20 MOX after three cycles, the average value is 36 gigawatt i
21 day per ton, and it the same of uranium. l 22 And the new type, the new core management 4 j 23 implemented today is the hybrid management with MOX in i
24 three bath and uranium in four batch, due to limitation on I
[0 l
g 25 burn-up of the MOX. The MOX reload with 16 MOX per cycle, NEAL R. GROSS COURT REPORTERS AND TRANSCRSERS 1323 RHODE ISt.AND AVE., N W.
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85 1 equivalence 3.25, and we reload 28 uranium assemblies with 2 enrichment of 3.7, and the burn-up for three cycle for -- !
3 the average burn-up of the MOX is 36 gigawatt day per ton, j 4 and.for four cycle for uranium the burn-up is about 44 5 gigawatt day per ton.
6 And the new core managements that we study
-7 today for the future and that we want to implement is real a four batch management for uranium and for MOX, and it 9 corresponds to 12 MOX equivalence, 3.7 percent, and 28 10 uranium, 3.7 percent, and with this core management, the 11 MOX average burn-up is about 45 gigawatt day per ton.
12 But today we must justify the rod behavior 13 with this burn-up, and it needs to improve the design of 14 the fuel rod, in particular, for the problem of fission l 15 gas release, the internal pressure of the rod.
16 The EdF experience. Today ten PWR are fueled 17 with MOX assemblies. I give the list on the slide. We 18- can indicate that very recently the tenth PWR is..'at
.19 Tricastin 1 in January, and we have a load in reactor, 584 3 20 MOX from the beginning.
l l
l 21 On the next slide you can see the map, the 22 France map, with in red the different plants with MOX.
23 The blue plants are the 900 megawatt plants, and the red 24 plants are the 1300 megawatt plants, and we have MOX in
! 25 Tricastin, Le Blayes, Saint Laurent, Don Pierre, and a
l NEAL R. GROSS COUFiT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.
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- _ _ 1
i 86 ;
t 1 Galie. I 2 Okay. Here we can see a histogram of the MOX !
I t
3 fuel load in our plant. With a change in the old design [
-4 for MOX from the begianing to 1995, and in 1995 we changed j 5 the design of the fuel, and we adopt a new design, FR2G i 6- like for uranium fuel. !
7 Our operation experience. About core physic 8 start-up tests, we consider that we have obtained a good 9 consic;tency bet. ween predicted value and measurement in l
-10 term of boron cencentration at start-up, in term of !
l 11 temperature coefficient, control rod banks efficiency, and 5
.12 we have also obtained a good consistency between predicted 1
- 13. value and measurement on the core measurement uranium
- 1
') 14 operational tect, in particular, flume-ups each month, and l
15 local repeating factor _for its employ in the analyze of l l !
16 flume-ups.
I 17 We consider that these different results give 18 a good validation of the safety assessment computation.
19 We have some detail results of two slides. It's a typical 20 map of the core with the repetition of fresh plutonium is
-21 in green. One cycle plutonium is pink, and two cycle i 22 plutonium in the red, and the fresh uranium is in the 23 yellow in the very third part.
5 24 In France we use today the core management "j i T i i j 25 out-in with just in.the bottom of the median for effluence 1
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l
87 '
1 consideration.' We don't use fresh '
t 2
This is a demonstration of the good accuracy 3 of.the calculations, the good comparison between ;
4 calculation and predicted value in term of. boron 5 concentration, temperature coefficient, and rod j 6 efficiency. -
1 i
7 And this is the ram present (phonetic) !
l
-8 comparison between MOX and uranium for the' calculated i
9 value and the -- it's the difference. Each histogram 10 presents a difference between calculation and predicted 11 value on the flume-ups. ;
i 12 Okay. -The experience in terms of operation'of !
i
), 13- the plant. It's important in France for each nuclear 14 power plant to realize a grade full operation (phonetic), j i
15 its frequency control, a standard ELPO operation, and we
! 16 test this type of operation during some years on the two 17 plants of Saint'Laurent B, B-1 and B-2, and we obtained 18 good feedback of this type of operation. We don't have 19 any problem of operation, and we observe that we have 20 reduced liquid waste volume generated, used to the less l 21 effective xenon, xenon efficiency, permit to reduce the 22 liquid waste volume,'and also permit to have better axial-l 23 control. !
24 'And we can indicate also that we have realized i
25 under STUDVICK the test loop power run on MOX-rod, and we i NEAL R. GROSS
, CGURT REPORTERS AND TRANSCRISERS j l 1323 RHODE ISLAND AVE., N.W.
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~ ~ - r m
. . . . . ~ . . - , .-- -.- -.. --..-.----. -- - - . .
t 88 l
1 observe that MOX rods have better behaviors than uranium 2 in terms of PCI, and we obtain in 1995 as a general :
3 regulatory position for all the plants with MOX of '
4 operation will follow (phonetic). !
]
5 About the fuel behavior, today we have loaded 6 ten reactors, 584 MOX fuel assemblies. We have unload 300 7 MOX fuel' assemblies after three cycles, and the average 8 burn-up, discharge burn-up is about 37.5 gigawatt day per 9 ton after three cycle. The maximum value is about 40 10 gigawatt day per ton.
11 1 We have also implement in 1992 a four cycle 12 for four fuel assembly, and we have reached the average i 13 burn-up for these four assemblies of 44.5 gigawatt day per
'O 14 ton. H 15 And today we~have a new experiment of one j i
16 assembly in the center of the core with an objective for 17 increased burn-up to obtain 48 gigawatt day per ton for 18 one assembly. It's implemented in the Gravlene (phonetic) 19 reactor.
20 on this slide you can see the histogram of the a 21 burn-up, discharge burn-up, and you can see the average of 22 37 and a maximum value about 40 and the experimental 23 assembly.
24 Next.
i
(~ 25 For 1987 to today, MOX assembly BFU has been NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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l 89 ;
s i satisfactory. We just have one failure on one assembly in f
.t 2 July 1993 in Don Pierre 1 at the end of cycle 11. It's
{
3 probably due to debris in'the bottom part of the fuel :
L 4 assembly, and the fuel assembly has been reloaded because we have a quit area'and we'are inside the quit area, and.
~
5
)
6 we can_ reload liquid assemblies if it holds very little. !
l 7' In conclusion, we' disposed today of nine years 8 of operation, 45 years of experience, and we have the same 'i L
' f 9 reliability than uranium.
i 10 It's this one. Excuse me.
l f.
! 11 Okay. The outlook for EdF, it's that I ;
12 explained in the beginning, to implement our strategies.
r 13 The first aspect, the first item of our strategies, it's !
l
[~)
\/ y i 14 the flow equality strategy. We want to implement this i l 15 strategy. Today we have ten nuclear power plants burning t
16 MOX fuel assembly. .
L 17 During the two next years, in '97 and '98, we' ;
18 consider that we'can implement MOX core management on 16 i 19 nuclear power plants. It's nuclear power plants with 20 existing plutonium recycling approval and to the year 2000 l 21 we want to implement this type of core management on over !
22 28 nuclear power plants, but we need to implement public 23 inquiries on the 12 new power plants, and to day under l 24 CHINON plant, the first public inquiries is in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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L _ -
90 1 1 And the second item is the achievement of 7- 2 uranium-MOX parity, and to implement this new type of core f
1 3 management for the MOX, we need to increase the ;
4 performance of the MOX, and we need to experiment and 5 study of higher discharge burn-up rate, and we dispose of 6 irradiated rod examination, and for example, the different 7 experience that I explained of four cycle assemblies.
8 It's needed a few rod optimizations to 9 accommodate greater efficient gas release. Fuel 10 management is to be optimized to obtain a better 11 economical optimization and interest, and we need also to 12 implement vessel safety studies because of the impact of 13 higher plutonium quantity on the control system.
] 14 Just before the conclusion, I want to say some 15 -- to give some information about the reactivity insertion 16 accident, in particular, on the CABRI tests on the MOX 17 fuel rod. We disposed today of two tests on MOX fuel.
18 The test, REP Na6 and the Test 7, and the recent test of 19 seven have give a failure on the rod at deposition energy 20 of 110 calorie per gram, and you have the errors. Those 21 are the different hypotheses of the two tests, with 22 particular the current six test on the rod before there 23 was a test and the local burn-up, the REP Na6. It's a rod 24 with three cycle, and the REP Na7 in a rod with four r ~ s.
( j 25 cycle.
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91 1 Can you give the next?
es 2 On this slide you can see the -- all the tests 7
(v) 3 realized on uranium fuel rod with the REP Na1 was failure 4 and all the other tests without failure, with a different 5 deposit energy during the CABRI test.
6 And perhaps to explain, you can pass the next 7 slide. It's a curve, yes. It's a curve. Within the 8 horizontal is burn-up and in vertical is deposit energy 9 during the transient or at the failure instant, and you 10 can see on this slide the two point of MOX is the cycle 11 Na6. It's a three cycle rod, and Na7 with failure at 110 12 calorie per gram.
p_ 13 And our analysis of this result is that this
( )
14 result confirm that for our core management with MOX 15 management we don't modify the safety analysis and we 16 don't have a problem of safety due to the fact that the 17 core management is significant of the burn-up of REP Na6, 18 like we have seen the maximum burn-up in this type of core 19 management is about 40 gigawatt day per ton for the fuel 20 assembly. It's about 45 to 47 for the LOCA rod, and the 21 calculation of stored energy during rod ejection give l
22 result of about 60, maximum of 60 calorie per gram in the I j
i 23 safety analysis.
1 24 And with very ordous (phonetic) methodology of '
l ,f
(;j 25 the safety analysis report, and you can see that if you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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92 1 limit on your' safety analysin the deposit energy to 60,
- 2 you have some margin to the point REP Na6, and for the l 1
3 point REP Na7, you have also margin in terms of deposit .
4 energy. You have also margin in burn-up.
5 Today the REP Na7'is a very.recent test, and -
^
I 6 you don't dispose of all the detail-of the analysis of- l 1
7 this test, but these studies are implemented today, and we i i
8 must wait for the detailed result. !
9 This slide presents the actual situation of' l
10 the PWR nuclear power-plant in France with the different )
11 core management implement. On the left side, the'CP0 is. }
-12 oldest. plant, _are in the four batch, 3.7 percent. The 28 l f
13 900 megawatt plant are 18 in four-batch, 3.7 percent, and l O 14- ten in hybrid MOX core management, and for the 1,300 l i
}
15 megawatt plant', 13 are in the three batch 3.1 percent core ,
'i(
16 management. It is the origin core' management, and today ;
17 we implement the new type of core management. It's four !
18 batch, four percent. It's ex standard land cycle. i i
19 And if we project in the future, in the year f
\
20 2000, we have the CPO, the' oldest plant, ex standard land ;
l 21 core management, three~ batch, 4.2 percent. On the PWR 900 l l
l 22 megawatt, the 28 plant, MOX -- four batch MOX core j i
23 management with MOX equivalent 3.7. On the 20 1,300 j l 24 megawatt plant, the three batch, four percent, the !
25 extended cycle land (phonetic) , and for the new plant, the l
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93 1 four new plant, the same core management, three batch and f3 2 four percent on each one.
3 Here it's a projection in the future of the 4 MOX fuel load in PWR, the same histogram to 1996 and after 5 projection, and you can see that we have a goal of about 6 100 down to 20 percent -- 120 ton of heavy metal AIR 7 during -- at the end of the century.
8 And here is accumulated value. It's the same 9 thing, but an accumulated value, and you observe that 10 today we have recycled 261 ton of MOX, and the objective 11 is to recycle about 800 ton of heavy metal in year 2000.
12 On my conclusion, reprocessing / recycling 13 technology has reached maturity in France, and EdF I ID
's 14 considers the operation of MOX recycling PWR satisfactory 15 today. MOX on uranium core behavior are roughly 16 equivalent in term of operation and in term of 17 reliability, and the EdF goals for the year 2000 is 18 plutonium flow equality strategy, to keep down the 19 quantity of separate plutonium, and it's the uranium-MoX 20 parity to reach the same performance between the two 21 products, in particular to obtain high burn-up of MOX.
22 And in conclusion, EdF has confidence in these 23 new fuels of MOX.
24 Thank you for attention.
(
(3) 25 MR. WALLACE: Thank you, Jean-Luc.
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-_..._ ._ -.. . .m..._ ._ . _ .. _ _ _ _
94 i 1 Our last' speaker this morning is Dieter Krebs. !
2 Dr. Krebs is Director of Fuel Research and Technology at O 3 Siemens Power Corporation in Richland, Washington, and l
l I
- 4. also for-Siemens Power Generation Group in Erlinger, !
i 5 Germany. He has 27 years of nuclear plant and fuel design f
i
'6 experience, and in particular, appropriate to today's ;
i 7 discussion, with Mox experience in the U.S. and Europe. 3 8 We welcome his summary of the BWR European -!
i 9 experience, f 10 DR. KREBS: Ladies and gentlemen, I have first 11 to apologize. I had to step in on short-notice ~for a !
12 German utility colleague, and therefore, my viewgraphs are l I
13 collected from what was available, and they are not.that '
.f 14 nice as my French colleague had just before.
-15 (Laughter.)
'16 DR. KREBS: First one, please.
17 It shows the operating experience with Siemens 18 Mox mixed oxide. fuel, and it shows on the.one' side the 19 early test phase where we started with BWRs, as well, in 20 Germany, the first three plants, and in the U.S., as the 21 former Ex Nuclear, which is now Siemens Power Corporation, 22 but in about the same time frame also started with PWR in 23 Germany and even with a heavy water reactor also in 24 Germany.
.( ) 25 Then on the right side:you see what we call NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHoDE ISLAND AVE., N W.
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95 1 the commercial MOX insertion phrase, which started with g- 2 PWRs and then at the bottom you see the start of the BWR
\y) 3 program. Those are two advanced boiling water reactors, 4 the two 1,300 megawatt electric Gundremmingen units.
5 The next viewgraph shows the licensing status i 6 of light water reactors in Germany. Due to our federal 7 structure and the situation that nuclear licensing in 8 Germany is a state issue, the licenses and the license 9 values per plant differ quite a lot.
10 The next viewgraph elaborates that a little 11 more on the boiling water reactor side, where the first 12 three lines show the licenses which are already approved 13 and are used, so they are of all three types. There are I \
\/ 14 fuel assemblies under irradiation. Then it shows for two 15 more plants what is under licensing review, and at the 16 bottom, again, for the Gundremmingen plants it shows a 17 more advanced MOX design.
18 Next slide, please.
19 This shows -- gives a rough overview of the 20 safety evaluation related to MOX fuel assembly, which is 21 done in the license review. It is done for normal 22 operation in the reactor core in the spent fuel pool and 23 in the new fuel storage area and for transients and 24 accidents, and it covers all the areas of analysis from 13
(_) 25 neutron physics, thermal hydraulics, systams dynamics, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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96 1 fuel rod design, fuel structure design, and LOCA analysis,
,r w 2 and deals, of course, with the radiological aspects.
i i V
3 The next viewgraph, please.
4 For normal operating conditions, the 5 parameters of main concern are linear heat generation, 6 critical power ratio, thermal hydraulic compatibility with 7 the uranium assemblies, and shutdown reactivity, and the 8 criteria are the same for the uranium and for MOX fuel.
9 The behavior of the assemblies are somewhat 10 different, as my colleagues already explained, but they 11 have to fulfill the same criteria.
12 And the same is true with regard to transients
,_ 13 and accidents. No matter what it is, and the main areas i I
'# 14 to be analyzed are LOCA, reactivity, transients, and MCPR, l i
i 15 and stability analysis on the boiling water reactor side, 16 and again, there are no new criteria. The criteria to be l I
17 met are exactly the same.
18 The behavior is slightly different. Sometimes 19 it goes in the nice direction. Sometimes it goes in the 20 somewhat more demanding reactions, but the overall 21 behavior is the same. It's those effects more or less 22 balance out.
i 23 Then next slide, please.
24 There are additional criteria. They are
,f~')
(. ,/ 25 listed here. I don't want to read them all, and again, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS
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97 1 there are no new criteria. So this is the licensing f~s 2 approach. There is nothing new. I v
3 Then a question which comes up, especially if l 1
4 one enters into such a field with license, is about the 5 data phase, and you all know the problem of the egg and l 6 the hen, what has to come first, and in Germany the l
7 licensing approach with regard to MOX data was the basic l
8 assumption that MOX behavior is similar to UO 2-9 of course, we have then to check the validity 10 of that assumption by a limited number of data points, and )
11 then there are two possibilities. Either this basic 1
12 assumption is proven so it is valid, and then we don't 1 1
13 need an extended MOX database, and one example -- and I r~T l
! )
\~/ 14 will in a moment show you scme data on that -- is ramp 15 behavior, and it was already mentioned before by my French l 16 colleague that there is not worth behavior, and the same )
17 holds true with fission gas release.
18 Of course, the fission gas release depends on 19 the power history of the rod, but if we look at same power 20 histories in MOX and in uranium, the data show that they l l
21 are equivalent. So we can deal with that in fuel 22 management if we insert the MOX assemblies in a way which 23 is not more demanding than for uranium. We have the same I
24 fission gas release if we insert them in a more demanding 1
(~h 1
() 25 case. Then we have to adapt for that, but fission plena l l NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.
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98 i
i can easily be adopted.
E
, 2 If it were the basic assumption not to move,
! 3 then of course there would be the need for an extended on-4 4 standing MOX database, but up to now we have not found 5 such cases.
6 Our. design goal in Germany for MOX insertion 7- is burn-up as we usually in not a very precise way term 8 it. It's burn-up equivalence. What it.really means is we 9 are looking for the same site length, the same number of 10 fuel assemblies.
11 This shows you some data on the comparison of 12 ramp behavior UO2 versus MOX. The circles are UO. 2 The 13 full circles are failures, and the triangles are MOX, and 14 we haven't produced any MOX failure in ramp testing.
15 It seems to be that MOX behaves more benign 16 with regard to ramp behavior since up to now we are only 17 dealing with 90X mixed cores, partially MOX, partially i
18 uranium, and are not yet in any full MOX core. There is I 19 no need to establish the limit for the MOX oince the 20 behavior of the mixed core is determined by the obviously--
21 somewhat' lower PCI threshold for the uranium. d 22 Next slide, please. This gives an overview of 23 all-the pool site inspections and hot cell examinations we l 24 did perform between about 1981 and mid of '96, and this is A
Q
~
25 just shown to give you the impression we are not doing j NEAL R. GROSS i COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W. ,
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99 l l
1 only a very small portion, although I said we to a large !
l -
2- extent also rely on the UO2 databases, but-the cross-check i
3 worth of this basic assumption is done, is really i i !
- 4- detailed. *
?
5 Next slide, please. .
. 6 What is.really different? One aspect which is .
- j. 7 different is the reactivity behavior, here shown as K
- j. 8 infinity, the multiplication factor over burn-up, and it ;
9 shows three cases. The dotted line, which is PWR'UO2 with
- j. 10 regard to BWR. The curves are somewhat different, but in ;
i 11 principle the difference between UO 2 , commercial MOX, and :
12 weapons grade MOX are similar. !
~
13 The dotted line shows the UO2 behavior. The j
'O. t V 14 thin solid line shows civil or reactor grade plutonium l i
15 behavior, burn-up equivalent in the sense as I explained 16 it before, and those two cases bound or envelope the case !
17 of weapons grade plutonium.
18 Next slide, please.
19 Here. you see some examples of typical ,
20 plutonium isotopic compositions. Wnen we. started in j
)
21 Germany, the early demonstration programs were all made ;
I i
22 with plutonium repossessed from British MAGNOX fuel, and 23 it had a pretty high fissile content of almost 79 percent 24 pool fis. (phonetic).
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100 1 reactors, the fissile content went down. Today what we
,- ) 2 receive now from reprocessing is in the order of ten I i V
3 percent less, 69. What we expect in future is in the 4 order of 66. What the lowest content we ever put back 5 into a reactor was from test second recycling, second 6 reprocessing of over time fuel, and that had as low as 57 7 percent pool fissile, and that in combination with the 8 previous figure tells you we have more or less covered 9 that variance of the fissile content and can easily 10 extrapolate into the isotopic vector for weapons grade 11 plutonium.
12 Next one, please.
13 This shows the design of the BWR fuel assembly y
\
\ s' 14 inserted right now in Gundremmingen. The advanced here 15 stands for its carrier material is tense (phonetic) 4 16 uranium. We have also under irradiation same type of fuel 17 assemblies, the same structural design with natural 18 uranium as a carrier material.
19 Next slide, please.
20 This is a former fuel assembly type, a more 21 advanced design, a nine-by-nine, the ATRIUM 9 assemblies, 22 and it shows on the right side the MOX design and on the 23 left side the corresponding uranium design, and as you can 24 see, they are pretty rauch the same.
r~T
() 25 The same holds true for the next more advanced NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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i 1 101 l
j 1 design,zthe ten-by-ten. Here'I don't show the cross-
{. 2 section again since this is, again, very similar. This i
- .s 3 shows the axial layout of the ATRIUM 10 MOX assembly, 1
{ 4 which is -- we hope to insert after completion of the 5- licensing. process next year in Gundremmingen, as the most a
l '
6 advanced BWR MOX. design, and it shows the axial layout of-1
! 7 the'MOX on the left side, and on the right side the i
- 8 uranium. '
9 There is a difference. We don't have.a-10 natural uranium blanket applied for in the MOX desi~gn.
-11 The pure reason is manufacturing cost. The more-4 12 complicated you make the neutronic design of MOX assembly i
- 13 in order to minimize peaking factors, the more you drive I\ 14 the manufacturing cost,'but of course, as we all know, you j 15 can design fuel management schemes with somewhat less c
16 neutronically optimized fuel assemblies, and here we '
i 17 looked into an overall optimum also with regard to 1
- 18 manufacturing cost.
l 19 Next one, please.
j 20 The next few viewgraphs are supposed to show s-l 21 you some differences in the core bshavior of boiling water
{ 22 reactor MOX, and it shows the void coefficient, solid line e
- 23 for UO 2 , and dotted line for a. mixed MOX core with almost f 24 40 percent of MOX for the most advanced design, the' ATRIUM j 25 10, and as you can see, there are some differences, but j
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102 1 they are not dramatic at all.
- 2 Next one, please.
[~h 3 This shows again for the same case, ten-by-ten 4 design, for 1,300 megawatt boiling water reactor the 5 behavior of the linear heat generation rate, and you see, 6 yes, there are some differences, but they are, again, just 7 minor, and they are minor over the whole burn-up range.
8 Next one, please.
9 Same holds true for the thermal hydraulic 10 behavior, the MCPR.
11 With regard to codes, of course, we have to, 12 as was said earlier, we have to benchmark them, and we did 13 that. This shows one example. For the neutron physics g/
(~- 14 code we used the same code system as we use for uranium 15 cores, the combination of CASMO and MICROBURN, and the 16 validation was done by comparisons with critical 17 experiments with gamma scans and with -- on an analytical 18 basis with Monte Carlo calculations, and of course, we 19 looked into the core tracking, code critical conditions, 20 and TIP management measurement data.
21 Thank you.
22 MR. WALLACE: Okay. Thanks very much, Dieter.
23 And like we did with the first half of the 24 morning session, we'd like to open it up to general r
(m)
+j 25 questions initially related to the two presentations we've NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 1 just heard, and members of the staff questions preferred fm 2 first.
L ,)
3 (No response.)
4 MR. WALLACE: Okay. From the general l
l 5 audience?
l 6 Obviously we've done such a great job in i 1
1 7 preparing and communicating we've answered the questions I 1
8 folks might otherwise have in mind.
9 Okay. I'll just generally say are there any I 10 questions at all related to the presentations, either the 11 second half or from the first part of the morning.
12 Okay.
13 MR. ARCHIBAL: This is Ralph Archibal from 7-.s
\' 14 NRR.
15 I'm just curious about your plans and 16 schedules for anything that would really be coming in for 17 review and whether NEI is going to be the focus point, et 18 cetera. Maybe more generic, but in the reactor area, are 19 there any near term plans for us for resource purposes and 20 planning or anything like that, expectations?
21 MR. WALLACE: A good question, Ralph. I'll 22 attempt to answer that in telling you that what we are 23 going to be doing is an overall working group is 24 continually looking to bring to a common focus those O 25 issues that can be addressed in a uniform way so as to
( ,/
l l NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS i 1323 RHODE ISLAND AVE., N W.
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104 1 facilitate your interacting with the industry and vice
,g 2 versa.
U 3 I think as our working group goes forward, we 4 have a subcommittee that's referred to as the NRC 5 Regulatory Group, specifically intending to identify those 6 issues that are going to require interaction and what the 7 level of resources might be and the timing and so forth.
8 I'd like to ask that outside of this meeting 9 that group should be able to engage the NRC staff in what 10 the specific plants might be as best we can see the 11 process going forward at this point. So we'll get back to 12 you on that.
13 Yes?
(~h
- 14 MR. LYMAN Hi. I'm Ed Lyman from Nuclear 15 Control Institute.
16 I was struck by your remark, the gentleman 17 from EDF, that fuel redesign might be necessary to l i
I 18 actually bring MOX fuel burn-ups up to around 45,000 l 1
19 megawatt days per ton. Now, is that fuel redesign work 20 going on now?
l 21 Since COGEMA is one of the proposed vendors 22 for MOX fuel from the United States, we want to be able to l
l 23 qualify fuel for burn-ups around 45,000. I mean, is there 24 going to be extensive fuel development work? Where in the
( ,) 25 process now is the development of that fuel?
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, 105 l
1 Because clearly the experience that you've 2 already had with the older fuel designs _is not going to be l
- l. 3 relevant. So how much more design work and development do l-4 you expect for these redesign fuel elements that would be 5 applicable to the U.S. situation?
l i
6 MR. WALLACE: Jean-Luc, I'11 ask if you might i
7 be able to respond to that question.
- 8 MR. PROVOST
- I can try.
t
> 9 Yes, the design of the MOX fuel that we need 10 today for our core management, the core management as I 11 explained, need the burn-up that I explained, the maximum i
l 12 burn-up of about 40 gigawatt day per ton, and today we j l 1 13 don't need for this type of management more, but we are 14 studying new core management and to increase the burn-up 15 to 45 or 50 gigawatt day per ton maximum assembly, i
16 And I don't explain here that it's not 17 possible, but we have not experience of -- large 18 experience of this burn-up. We just have some assembly at 19 45 megawatt day per tons that I explained. It's in four 20 assembly, and today we have one assembly which be unloaded 21 at 48 megawatt day per ton, and we hope that we can obtain 22 more feedback on this type of fuel to the future, to l: 23 increase the performance of this fuel.
l 24 But like I explained, perhaps it needs to
' 25 increase the performance of this fuel to make modification I
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i 106 1 of the rod, of the fuel rod, but I cannot explain more in g 2 detail today because it's in studies, the fuel supplier j 3 study, some improvement to increase the burn-up concerning i
4 this problem of efficient gas release. !
i l5 That is the same as the German? !
l 6 DR. KREBS: Could I also comment a little bit 7 on that?
8 We have in Germany experience with as well on j 9 the UO2 as the MOX area on the PWR side up to almost 50
- 10. megawatt days per kilogram, and we have not seen any I 11 difference, any.significant difference between UO2 and MOX. l 1
i 12 of course, whatever we look at, be it uranium, 13 be it MOX, if we design for increased burn-up, we have to 14 look at'the details, but those'are minor evolutionary 15 changes. If we do it on the uranium side, the two main 16 issues we have to look in is cladding, water side cladding-17 corrosion and its fission gas release, and it's exactly 18 the same on the MOX side.
i 19 So this is nothing new in that-area. It's the 20 normal evolutionary process for increased burn-up.
l
! 21- MR. WALLACE: I just might add we appreciate 22 the collaboration of our colleagues from Europe-because it l
23 helps us to take account of the very practical experience 24 they are having now in an area where we don't have similar L ~3 25 experience. l l
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I 107 1 Yes.
7 2 MS. ARRIGO: I'm with Nuclear Information and
\,
~~'
)
3 Resource Service.
4 I think -- I couldn't see the chart real well, 5 but the Electricite de France chart showed an isotopic 6 composition of plutoniums in MOX or maybe there were other 7 isotopes. I couldn't see.
8 My question is regarding the amount of fission 9 products and the comparison between those in MOX and in l 10 uranium oxide fuel. What kind of studies are available to l 11 compare those, if you have those?
l 12 MR. WALLACE: If I understand your question, 1
13 you're asking what the comparative difference is between 75 i l
I V 14 low enriched uranium fuel and MOX fuel from a point of 15 view of fission products in the spent fuel? l l
16 MS. ARRIGO: Yeah.
i 17 MR. WALLACE: Okay. Jean-Luc? I 15 MR. PROVOST: It's a difficult question for l l
19 me.
20 For what I expect you ask this question, for i 21 the problem of does this under spent fuel or for -- ]
22 MR. WALLACE: Spent fuel. l 1
23 MR. PROVOST: For the spent fuel? '
24 MR. WALLACE: Any differences in spent for MOX
( ) 25 versus --
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.108 1 MR. PROVOST: Yes. I give some information I
- 2 about the spent fuel and the spent fuel transport, and we
, I
'3 -- perhaps these problems that you explain The problem )
1 4 is that we.must have more important cooling time on the 5 MOX fuel than on the uranium fuel. The main aspect is the 6 presence of Plutonium'238 in more quantity.under spent MOX 7 fuels than under spent uranium fuel, which give calorie, 8 temperature and they give radiation, but not only this L 9_. aspect. There is also the different fission produced, but l 10 I don't know exactly the difference.
11 Perhaps people can answer.
12 MS. ARRIGO: Then I was going to ask if you 13 know of who would be looking into that, where I could get i
,b l 14 more information on that if you don't have that. l l.
l 15 MR. WALLACE: I think we may need to get back l l
16 to you on that. I think the essence of Jean-Luc's l -17 response is there are differences that they recognize, and !
18 that's taken into account in what they do, and beyond that i
19 I don't think we're prepared to say any more at this time.
l.
20 PARTICIPANT: Sherman (inaudible) of OGC of R l-21 NRC.
22 'Just a general question. What percentage of
-23 your countries use nuclear fuel for the electricity j 24 production, and then what percentage of those plants are 25 using MOX fuel?
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~. . , - , -
- . . . - - . . ~. .-. -_- - . . _ ,-
109 1 MR. PROVOST: In France, the nuclear
,y 2 electricity is about 82 percent of the electricity (v )
3 production, .and I can explain that there is ten reactor 4 which use MOX fuel for 55 reactor in PWR reactors. It's 5 20 percent of the reactors which use MOX today.
6 But the objective at the end of the century is 7 about half of the PWR which use MOX.
8 DR. KREBS: In Germany, the nuclear generated 9 electricity is in the order of one third of the national 10 grid, and up to now about a little more than half of the 11 nuclear power plants have already gained some experience, 12 and who has got a license you saw in one of the 13 viewgraphs, and since in the past all of the -- due to the
[, ')
\ '# 14 German atomic law, all of the utilities went into i 15 reprocessing only after the recent change. There is now 16 the dual way of either reprocessing or direct storage. So 17 all have some plutonium to burn.
18 Some are considering to combine it so that 19 it's not yet really clear how many reactors will do it, 20 but at least 50 percent of the German reactors are going 21 eventually to have mixed cores.
22 MR. WALLACE: Any other questions?
23 (:No response.)
24 MR. WALLACE: Okay. I just have a few closing
( ,) 25 comments that I'll make by way of summarizing a part of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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110 1 what we've said, and I'm going to use the mission slide, rs 2 the one in the front of your package, just to make a
( )
3 couple of points.
4 One, the direction in which the United States 5 may eventually go with respect to nonproliferation and 6 weapons grade plutonium disposition is a decision that 7 will be made by the policy makers in our government. The 8 role that we in the industry have is to be prepared to 9 implement that policy statement when and if it's made.
10 Points 3 and 4 on there in that regard are 11 particularly relevant and the basis for the interactions 12 that we have had today and which we will have going 13 forward.
(x -) 14 It is our purpose through the NEI working 15 group to resolve generic regulatory and legal issues and 16 bring an efficiency and an effectiveness to the process in 17 so doing that.
18 It also is our objective to facilitate 19 communications within the industry and between the 20 industry in the various governmental bodies that are 21 concerned, and in fact, as I mentioned at the outset, the 22 make-up of our working group truly makes it the industry 23 that's involved in this activity, with 11 utilities 24 representing the significant utility operators in the p.
() 25 country, and virtually all of the other vendors, NEAL R. GROSS COURT REPORTERS AND TRANSCRIGERS 1323 RHODE ISLAND AVE., N W.
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l l
111 1 suppliers, fabricators who have a part of this process.
rx 2 This working group really does represent a coalition and a 3 collection of the talent of the industry.
4 The group, I can tell you, works with a sense 5 of cooperation and collaboration, and it's in that light 6 that we would expect to be able to work with the NRC and 7 others in order to carry out the policy directions that 8 the government may in the U.S. here ultimately take.
9 What we have done here today has a particular 10 implication of what we are carrying out, those last two 11 mission statements.
12 Further, the next meeting on March 26th, which 13 will have a focus particularly on fabrication, will be an
()
-I 14 extension of that, and we would expect future 15 interactions, as I indicated earlier in response to a 16 question, with the NRC to also bring forward a common way 17 of communicating to facilitate this program being 18 developed in the future.
19 In summary of what we've heard today, I think l
20 the overall conclusions are perhaps obvious and don't need 21 to be stated, but the U.S. view, as discussed by several l 22 of our vendors, and the European experience indicates that 23 MOX fuel can be used in U.S. light water reactors, and 24 moreover, that it will perform in ways that are not very C
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112 1 light water reactors.
,3 2 With that, I thank all of you for attending.
i s kj 3 The presenters will be available at the front of the room 4 should there be additional questions, and we look forward 5 to uur next open technical meeting on March 26th, hosted 6 by the NRC.
7 Thank you.
8 (Whereupon, at 12:31 p.m., the meeting was 9 concluded.)
10 11 12 13 4
(~x)
'u/
14 15 l l
16 17 18 a
19 I I
20 21 22 23 24
/'"T l g 25 I i
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. - . _ - - . = - . . ..
1 A
L)
CERTIFICATE This is to certify that the attached proceedings before the United States Nuclear i
Regulatory Commission in the matter of: )
Name of Proceeding: INDUSTRY PRESENTATION ON THE USE OF i MIXED OXIDE FUEL: OPEN TECHNICAL MEETING !
i Docket Number: N/A l 4
I
+
Place of Proceeding: ROCKVILLE, MARYLAND were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear l/O
, e) Regulatory Commission taken by me and, thereafter reduced to
] typewriting by me or under the direction of the court reporting company, and that the transcript is a true and 1
accurate record of the foregoing proceedings. l LMj W A n a-+
(4.AURIE ANDREWS Official Reporter Neal R. Gross and Co., Inc.
1
+
k NEAL R.. GROSS COURT REPORTERS ANDTRANSCRIBERS 1323 RHODEISIAND AVENUE,NW (202)234 4133 WASHINGTON, D.C. 20005 (202)234-4433
.