ML21036A180
| ML21036A180 | |
| Person / Time | |
|---|---|
| Issue date: | 12/01/2020 |
| From: | Derek Widmayer Advisory Committee on Reactor Safeguards |
| To: | |
| Widmayer, D, ACRS | |
| References | |
| NRC-1260 | |
| Download: ML21036A180 (104) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Docket Number:
(n/a)
Location:
teleconference Date:
Tuesday, December 1, 2020 Work Order No.:
NRC-1260 Pages 1-74 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
(ACRS) 5
+ + + + +
6 FUTURE PLANT DESIGNS SUBCOMMITTEE 7
+ + + + +
8 TUESDAY 9
DECEMBER 1, 2020 10
+ + + + +
11 The Subcommittee met via Video-12 teleconference, at 9:30 a.m. EST, Dennis Bley, 13 Chairman, presiding.
14 15 COMMITTEE MEMBERS:
16 DENNIS BLEY, Chairman 17 RONALD G. BALLINGER, Member 18 CHARLES H. BROWN, JR. Member 19 VESNA B. DIMITRIJEVIC, Member 20 WALTER L. KIRCHNER, Member-at-large 21 JOSE MARCH-LEUBA, Member 22 DAVID A. PETTI, Member 23 JOY L. REMPE, Member 24 MATTHEW W. SUNSERI, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 ACRS CONSULTANT:
STEVE SCHULTZ 3
4 DESIGNATED FEDERAL OFFICIAL:
5 KENT HOWARD 6
8 ALSO PRESENT:
9 DON ALGAMA, RES 10 DREW BARTO, NMSS 11 AMY CUBBAGE, NRR 12 RICHARD LEE, RES 13 SCOTT MOORE, Executive Director, ACRS 14 KIM WEBBER, RES 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 CONTENTS 1
2 Opening Remarks 4
3 Staff Introduction 4
4 Non-LWR Code Development, Volume 5, Radionuclide 5
Characterization, Criticality, Shielding, and 6
Transport in the Nuclear Fuel Cycle 7
Discussion
................. 62 8
Public Comment (none)
............ 68 9
Adjourn
................... 74 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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4 P R O C E E D I N G S 1
9:30 a.m.
2 CHAIR BLEY: Good morning. This meeting 3
will now come to order. It's a meeting of the 4
Advisory Committee on Reactor Safeguards Subcommittee 5
on Future Plant Designs.
6 I'm Dennis Bley, Chairman of the Future 7
Plant Designs Subcommittee.
ACRS members in 8
attendance are Ron Ballinger, Charlie Brown, Vesna 9
Dimitrijevic, Walt Kirchner, Jose March-Leuba, Dave 10 Petti, Joy Rempe and Matt Sunseri will be joining us 11 in about an hour. And our consultant Mike Corradini 12 is in attendance for part of the meeting this morning.
13 Derek Widmayer of the ACRS staff is the 14 designated federal official for this meeting. Kent 15 Howard is the backup DFO for the meeting.
16 The purpose of today's meeting is to 17 review the draft NUREG Document NRC-Non-Light Water 18 Reactor Vision and Strategy, Volume 5, Radionuclide 19 Characterization Criticality, Shielding and Transport 20 in the Nuclear Fuel Cycle.
21 It's the final volume of the staff's 22 documentation of their near-term implementation action 23 plan for Strategy 2, computer codes.
24 The subcommittee will gather information, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 analyze the relevant issues and facts and formulate 1
proposed positions and actions as appropriate. This 2
matter will be brought to the February 2021 full 3
committee meeting along with Volume 4 of the NUREG 4
series for a possible letter report.
5 Previously on November 4 of 2019, we sent 6
a letter report to the Chairman of the NRC from 7
Volumes 1, 2 and 3 in an overview report. At the end 8
of the today's subcommittee meeting, the members of 9
the subcommittee and the staff will discuss plans for 10 the February 2021 full committee meeting.
11 ACRS was established by statute and is 12 governed by the Federal Advisory Committee Act, FACA.
13 The committee can only speak through its published 14 letter reports.
15 We can hold meetings to gather information 16 and perform preparatory work that will support our 17 deliberations at a full committee meeting. The rules 18 for participation in ACRS meetings including today's 19 were announced in the Federal Register on June 13 of 20 2019.
21 The ACRS Section of the U.S. NRC public 22 website provides our charter, finalized agenda, letter 23 reports and full transcripts of all full and 24 subcommittee meetings, including the slides to be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 presented here.
1 The meeting notice and agenda for this 2
meeting were posted there. And as stated in the 3
Federal Register notice and in the public meeting 4
notice posted to the website, members of the public 5
who desire to provide written or oral comments to the 6
subcommittee may do so and should contact the 7
designated federal official five days prior to the 8
meeting as practicable.
9 Today's meeting is open to public 10 attendance, and we have received no written statements 11 or requests to make oral statements.
12 We have also set aside 10 minutes in the 13 agenda for spontaneous comments from members of the 14 public attending or listening to our meetings. Due to 15 the COVID pandemic, today's meeting is being held over 16 Microsoft Teams for the ACRS and NRC staff attendees.
17 There is also a telephone bridge line 18 allowing participation of the public over the phone.
19 A transcript of today's meeting is being 20 kept. Therefore, we request that meeting participants 21 on the bridge line identify themselves when they're 22 asked to speak and to speak with sufficient clarify 23 and volume so that they can be readily heard.
24 At this time I ask that attendees on Teams 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 and on the bridge line keep their devices on mute to 1
minimize disruptions and unmute only when speaking.
2 We will now proceed with the meeting. And 3
I call on Kim Webber, Deputy Director of the Division 4
of Systems Analysis in the Office of Research to 5
begin. Kim?
6 MS.
WEBBER:
Yes.
Good
- morning, 7
everybody. I hope you all had a nice Thanksgiving.
8 I know that I am still eating turkey. And I've been 9
eating it since last Sunday, so I'm getting tired of 10 eating leftovers. But anyway, hope you all had an 11 enjoyable holiday and with that I'll get started on my 12 presentation.
13 First, I want to thank you for taking the 14 time to review our latest volume on code application 15 activities.
It's Volume 5,
Radionuclide 16 Characterization, Criticality, Shielding and Transport 17 in a Nuclear Fuel Cycle.
18 My name is Kim Webber. I'm the Deputy 19 Director of the Division of Systems Analysis in the 20 Office of Nuclear Regulatory Research. And we will be 21 asking for a letter on both Volumes 4 and 5.
22 Volume 4, you may recall, we presented to 23 you, I think it was last month. And so I think we're 24 also anticipating a full committee meeting sometime in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 the late winter time frame, maybe February or March.
1 Next slide, please. Okay. So with me 2
today are Don Algama, he's the Senior Reactor Systems 3
Engineer in the Office of Research, and Andrew Barto, 4
a Senior Nuclear Engineer in the Office of Nuclear 5
Material Safety and Safeguards.
6 They've been working very hard over the 7
last several months to develop a strategy that we 8
believe is the best approach to enable our readiness 9
to support safety reviews of the front and back end of 10 the fuel cycle.
11 Over the next few minutes, I'll provide an 12 overview of the status of the non-light water reactor 13 code development project and a short overview of 14 Volume 5.
15 Then I'll turn the presentation over to 16 Don and Drew, who are going to discuss the details of 17 Volume 5, including the topics shown on this slide and 18 in the agenda.
19 Could I have the next slide, please?
20 RES's mission now more than ever is to enable the 21 regulatory offices, like NRR, to be ready to perform 22 licensing reviews and oversight responsibilities for 23 advanced non-light water reactor technologies.
24 With that be ready attitude, we're doing 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 research differently, embarking on more be ready 1
strategies.
2 To improve mission value, we're working 3
hard to deliver the tools, expertise and information 4
in a cost effective and efficient manner so that 5
licensing can be completed on time and within the 6
allotted resources.
7 A key element of this strategy, as you 8
know, is developing the codes and analytical tools.
9 Direct code development activities and collaborations 10 with many organizations you see on this slide were 11 gaining knowledge and building staff expertise and 12 analytical capabilities to support safety analysis for 13 a wide range of advance reactor designs.
14 Next slide, please. To facilitate the 15 Agency's readiness, the NRC's near-term implementation 16 action plan was developed in 2017. The IAP is the 17 vehicle to execute the NRC's vision to safely achieve 18 effective and efficient non-light water reactor 19 mission readiness.
20 As you know, the IAP includes six 21 strategies and Strategy 2 focuses on computer codes 22 and knowledge to perform regulatory reviews.
23 Next slide, please.
24 MS. REMPE: Kim?
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10 MS. WEBBER: Yes?
1 MS. REMPE: This is Joy.
2 MS. WEBBER: Hi, Joy. Good morning.
3 MS. REMPE: Good morning. I had a 4
question, and I couldn't decide whether to ask later 5
or to ask you. But I think it pertains more than to 6
just Volume 5 so I think I'm going to ask you.
7 In our biennial report last time we issued 8
it, we recommended that RES review and update as 9
needed the Agency's non-LWR implementation action 10 plans to ensure that they emphasize the data that 11 design developers have to obtain to validate codes for 12 various new concepts.
13 And in the back of Volume 5, or I guess 14 actually it's on Page 13, there are some statements 15 that talk about the designs haven't provided enough 16 detailed information on non-LWR fuel cycle 17 implementations and so they realize that what they're 18 doing may have to be updated.
19 But we observed the need for updates 20 because when we started this non-LWR activity, there 21 were very few details and the designs have evolved.
22 And have you guys talked about when you think you're 23 going to be updating some of these plans, how often 24 they need to be updated? Or what's the trigger for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 trying to go back through and say what's still 1
applicable and what's not applicable or what else 2
needs to be added?
3 MS. WEBBER: Well, thank you for the 4
question. So generally our strategy involves 5
developing what we call reference plant models. And 6
so those reference plant models are based on publicly 7
available information of advance reactor designs that 8
are very similar to the ones that, you know, we 9
anticipate receiving.
10 So, for example, heat pipe reactors, we 11 have a reference plant model for heat pipe reactors, 12 sodium fast reactors, high temperature gas reactors, 13 et cetera.
14 And those reference plant models are being 15 developed not only in the context of the safety 16 analysis work of Volume 1, but they're being developed 17 in the context of Volume 3.
18 And the whole purpose for taking that 19 approach is to minimize the amount of time that it 20 would take to update the codes for design specific 21 information.
22 And so the plan really is that these 23 reports represent the global strategy and identify the 24 gaps that exist and the verification validation needs, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 et cetera. But that really, when it comes to doing 1
the design specific work, we're going to rely on our 2
existing user need requests and RER research assistant 3
request processes to, you know, do the more design 4
specific licensing work.
5 So that activity will not be incorporated 6
into any revision of these volumes. Does that help 7
answer the question?
8 MS. REMPE: Yes. But so let me rephrase 9
in a way to make sure I understand.
10 MS. WEBBER: Sure.
11 MS. REMPE: I was aware of the reference 12 plant evaluations. And so you're going to use that to 13 ensure that these volumes are sort of applicable.
14 That you're not going to ever update these volumes 15 because you will rely on what you learned from the 16 reference plan evaluations and design specific 17 activities to see if there are any gaps, and you'll 18 deal with it elsewhere. But it sounds to me like you 19 will not be updating these volumes. Is that a good 20 conclusion from your response?
21 MS. WEBBER: Yes. I would characterize it 22 slightly differently. So while these volumes 23 represent what we know to be the gaps today and the 24 verification/validation needs and the code development 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 tasks, you know, they were developed at a point in 1
time. And I would anticipate that unless there's a 2
substantial change relative to the information that's 3
contained in them that we will not need to update 4
these volumes.
5 But like I said, if there is a substantial 6
change, then one way to communicate our plans to 7
reflect that substantial change would be to update 8
whatever volume is needed.
9 MS. REMPE: Okay. So the reference plan 10 evaluations may identify the need for a substantial 11 change, et cetera, or some new design that you have to 12 deal with may identify the need for a substantial 13 change. But that would be the only reason that such 14 a substantial change would occur.
15 MS. WEBBER: Yes. Like none of these 16 volumes address fusion reactors, you know. And so 17 there are things that are probably out there a little 18 bit farther that when we started this work we did not 19 envision like fusion technology.
20 And so, you know, if that becomes a 21 reality then we'll have to start, you know, thinking 22 a little bit more deliberately about how we address 23 the gaps and the needs relative to, for example, 24 fusion technology.
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14 MS. REMPE: Okay. This helps. Thank you.
1 MS. WEBBER: You're welcome. Thank you.
2 CHAIR BLEY: Kim --
3 MS. WEBBER: Yes.
4 (Simultaneous speaking.)
5 CHAIR BLEY: -- just a little further 6
there. First, I would like to thank you for this slide 7
with the hot links to your updated volumes.
8 MS. WEBBER: Oh, good.
9 CHAIR BLEY: And I don't know if anybody 10 has done that before so I appreciate it.
11 MS. WEBBER: Well, I've got to thank my 12 staff for doing that.
13 CHAIR BLEY: Well, the introduction, 14 Volume 1, Volume 2, Volume 3, were issued in these 15 versions in January. I haven't been through those 16 yet. But are they updates of the ones we reviewed a 17 year ago?
18 MS. WEBBER: Well, so you may recall that 19 you -- I'm getting a weird echo. You may recall that 20 we issued the introduction, Volume 1, 2 and 3 and had 21 a meeting with you last November of 2019, I believe.
22 And then we updated these volumes to reflect comments 23 and feedback that we received through the various 24 meetings and also as a result of that letter.
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15 And so the versions that you see for the 1
introduction, Volume 1, 2 and 3 are the final set that 2
reflect modifications, the feedback that we received 3
from you. Now Volume 4, we had the subcommittee 4
meeting in, I think it was October.
5 CHAIR BLEY: Late September, but go ahead.
6 MS. WEBBER: Yes, late September. So this 7
one is still a draft. And the staff, I know that they 8
recently looked at the transcript. And so they're 9
trying to update that volume, you know, as we speak.
10 And then if we go into the full committee meeting, 11 they'll take whatever feedback from that.
12 And Volume 4 and 5 together, we will 13 finalize in a version that's, you know, sort of the 14 official Version 0 or Version 1.
15 So, you know, if you could see these 16 pictures on Slide 5 for the different volumes, you 17 would note that there's a date in there of, I think 18 it's January.
19 CHAIR BLEY: That's right.
20 MS. WEBBER: Yes. It's January. And so 21 that represents sort of the final Version 1 of these 22 documents, at least at this point.
23 CHAIR BLEY: So Volume 4, well, I guess 24 looking through the slides that the gentlemen are 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 going to provide next, it looks like you made some 1
presentations on kind of changes since Volume 5 was 2
published.
3 Do you expect you will revise to any 4
extent Volumes 4 and 5 before our February meeting?
5 MS. WEBBER: Well, we probably will make 6
some revisions. And, you know, if you're interested 7
in seeing, like, a red line strike out version of the 8
two volumes before the full committee meeting, we 9
would be happy to provide that if that would --
10 CHAIR BLEY: Thanks. That would be very 11 helpful. We would appreciate that.
12 MS. WEBBER: Okay. Yes. We could do 13 that.
14 CHAIR BLEY: Okay. One last question in 15 this area, and we won't talk about it at the end of 16 the meeting. The introduction was pretty thin when we 17 saw it the last
- time, and we noticed some 18 inconsistencies in approach in Volumes 1, 2 and 3.
19 Were those addressed and should we -- at 20 the February meeting, would it be worth 15 minutes to 21 half an hour to bring us up to date on what you 22 changed in introduction, 1, 2 and 3?
23 MS. WEBBER: Yes. We could do that. You 24 know, maybe we need to talk offline about the specific 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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17 interests that you have because I'm not clear on the 1
specific interests relative to doing that.
2 CHAIR BLEY: Okay. Well I'll have Derek 3
work with you and set up something to talk about that 4
because that might affect how we decide to write the 5
letter come February.
Sorry for all the 6
interruptions. Go ahead.
7 MR. PETTI: I had a question. This is 8
Dave. Since we're talking about the big picture here.
9 MS. WEBBER: Mm-hmm.
10 MR. PETTI: I think it's hard to write 11 Volume 5 so I don't want this to come across as 12 critical.
13 MS. WEBBER: Mm-hmm.
14 MR. PETTI: But I'm trying to understand 15 the backdrop here. You guys are envisioning, for 16 instance, fuel fabrication facilities and doing 17 criticality analysis of new fuel fabrication 18 facilities for advance reactors, which have different 19 fuels and LWRs. That seems to be something well 20 downstream in the future --
21 MS. WEBBER: Mm-hmm.
22 MR. PETTI: -- compared to said Volumes 1, 23 2 and 3 where, you know, the first reactor you're 24 going to do something with. The document is silent on 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 the fact that the first cause for these reactors are 1
probably going to come from down blended HEU.
2 MS. WEBBER: Mm-hmm.
3 MR. PETTI: It would have been made by DOE 4
or by commercial vendors that have a license to handle 5
HEU and HALEU. And so it kind of just, it threw me.
6 It would seem to me that a footnote or a paragraph 7
that recognizes where we are today relative to sort of 8
where you are envisioning it, you know, in a full, you 9
know, commercial setting --
10 MS. WEBBER: Sure.
11 MEMBER PETTI: -- where you've actually 12 got more than one would probably help because, you 13 know --
14 MS. WEBBER: Okay.
15 MEMBER PETTI: -- I mean, I didn't hear 16 anybody is much more focused on, you know, I need a --
17 I need HALEU now and that's a whole different 18 conversation. And then you read this, and it just 19 struck that you guys know this but the document 20 doesn't talk about that. And it makes it seem a 21 little, like, you know, out in left field.
22 MS. WEBBER: I think that's a good 23 comment. I think that's a good comment. And I 24 appreciate you for bringing that up. And that's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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19 probably something that we can address in the revision 1
to the report.
2 MR. PETTI: Okay. Yes. Because there's 3
a number of things like that where just footnotes 4
probably would help to just clarify some things --
5 I'll go through others as the slides come along so.
6 MS. WEBBER: Okay.
7 MEMBER PETTI: Thanks.
8 CHAIR BLEY: Yes, this is Dennis. One 9
last time, Kim. What Dave brought up resonated with 10 something that I've been thinking about. And this is 11 no surprise because these are delving into, in some 12 cases, into new areas.
13 It seems like the 10 reports you're going 14 to tell us about that are coming out of this plan --
15 this is substantially different than especially 16 Volumes 1, 2 and 3.
17 MS. WEBBER: Yes.
18 CHAIR BLEY: This is a plan, and those 10 19 reports are going to eventually get us to the kind of 20 evaluation you did for the other codes in 1, 2 and 3.
21 Is that correct?
22 MS. WEBBER: Yes. Yes, conceptually, I 23 think that's what's going to happen.
24 CHAIR BLEY: Yes.
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20 MS. WEBBER: But I think due to the 1
complexity of the fuel cycles for each of the 2
different designs and all the subtleties and nuances, 3
you know, I think the staff has done a really good job 4
of at least identifying, you know, the strategy in the 5
Volume 5 report.
6 And so, you know, once the strategy has 7
been identified, then I think they can focus more 8
specifically on a particular fuel cycle of interest or 9
a different step. And I'm kind of jumping ahead into, 10 you know, Don and Drew's presentation. But I think 11 it's at least a good start at a strategy to figure out 12 how best to do this.
13 And as, you know, you probably are aware, 14 a lot of the information on the fuel cycles is still, 15 you know, to be determined. And so we're really kind 16 of leaning forward to do the best that we can to 17 figure out what our information needs are and our, you 18 know, model development needs are.
19 And so, you know, this particular volume 20 is likely, you know, to evolve over time or the 21 strategies. And, you know, Don and Drew will talk 22 about it. But, you know, we're going to have to 23 prioritize based on, you know, what we see as the most 24 important steps.
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21 And we're already getting indicators, you 1
know, that people want to ship fuel for these designs.
2 I just heard the other day about someone being 3
interested in designing a package to ship fresh TRISO 4
fuel. You know, so the activities are already being 5
thought about.
6 You know, there's regulatory efforts that 7
are underway. And I'm pretty sure Drew can answer 8
some of the more detailed questions you might have 9
during his presentation. But if it's okay with you, 10 you know, let me just finish up my next few slides and 11 then we'll get into the details of Volume 5.
12 CHAIR BLEY: Okay. Go ahead.
13 MS. WEBBER: Okay. One last comment, I 14 think, Dennis, you had the question about, you know, 15 what the full committee meeting and the letter will 16 focus on. So we do have a letter on the introduction, 17 Volume 1, 2 and 3 and what we're seeking more 18 specifically is a letter on Volume 4 and 5.
19 So originally the thought was not to 20 necessarily go back and do a reassessment of the 21 intro, Volume 1, 2 and 3, but it was to really focus 22 on Volume 4 and 5. So that was at least my initial 23 thought, but we can talk about that.
24 So I think I've touched on, you know, the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 information relative to this slide. I guess the only 1
other thing that I wanted to point out for those in 2
the audience who may not have as much familiarity is 3
that, you know, Volumes 1 through 3 focus on the 4
systems analysis, fuel performance, neutronic source 5
term, severe accident progression and accident 6
consequence codes.
7 And then Volume 4
describes code 8
development plans for our suite of codes used to 9
evaluate the siting
- criteria, control room 10 inhabitability and other safety evaluations during 11 licensing. And then we talked about sort of the focus 12 for Volume 5 so we'll go to the next slide.
13 So, you know, if you'd like to follow the 14 status of our code development activities, you can go 15 to the advance reactor on our see public web page, 16 which is shown at the top left corner of this slide.
17 And then if you scroll down to the page 18 and then click on the summary of integrated schedule 19 and regulatory activities image, which is shown in the 20 bottom right-hand of this slide, then you'll see the 21 status of the major milestones for the near-term code 22 development tasks.
23 And a large portion of what we're doing 24 for Volumes 1 and 3 are these reference plant models 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 and building them out. And I think, you know, the 1
plans are to have much of that reference plant model 2
work done this year in 2021.
3 So can we go on to the last slide in my 4
presentation? So Volume 5 describes the staff's plans 5
to evaluate the ability of scale in MELCOR to support 6
safety analysis and licensing for front end and back 7
end of the fuel cycle.
8 By considering the fuel cycles for many 9
non-light water reactor designs, the staff developed 10 an approach that involves evaluating information gaps 11 and identifying methods that can be used to address 12 the gaps.
13 Using the light water reactor fuel cycle 14 as a reference point, the staff plans to develop a 15 series of individual reports, which we had been 16 talking about, and publicly available input decks that 17 characterize the co-development needs for all aspects 18 of fuel fabrication, transportation and storage as we 19 know them.
20 And, you know, due to the dynamic nature 21 of not only the advance reactor industry in terms of 22 designing their reactors, but there's also an 23 extremely dynamic fuel cycle process for each one of 24 those plant designs as well.
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24 And so now unless there are any questions, 1
I'll turn the presentation over to Don.
2 MR. CORRADINI: So, Kim, this is Michael 3
Corradini.
4 MS. WEBBER: Yes. Hi, Mike.
5 MR. CORRADINI: Hi, how are you?
6 MS. WEBBER: Good.
7 MR. CORRADINI: I hope you had a nice 8
holiday.
9 MS. WEBBER: Yes, it was great.
10 MR. CORRADINI: My big picture conclusion 11 from reading the volume and looking at your slides is 12 that the basis will be no core max and the current 13 tool scale.
14 MS. WEBBER: Yes.
15 MR. CORRADINI: And there will be slight 16 modifications as needed, but the overall structure is 17 already in place.
18 MS. WEBBER: Yes.
19 MR. ALGAMA: Yes.
20 MR. CORRADINI: Okay.
21 MS. WEBBER: And Don can talk -- I think 22 Don and/or Drew may talk more about that, Mike.
23 MR. CORRADINI: All right. I'm sure. I 24 just wanted to make kind of the 40,000 foot conclusion 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 is clear because I think that's personally the way to 1
go. Some of the suggestions Dave made might be 2
appropriate given the fact where the initial fuel 3
loadings will come from. But, okay. Thank you very 4
much.
5 MS. WEBBER: Yes. So just to expand on 6
that a little bit, as you know in Volume 1, we're 7
using new codes, Department of Energy funded codes.
8 But like Volume 3, we're going to use our own, you 9
know, well-known codes and filling gaps wherever those 10 gaps may exist. All right.
11 MR. CORRADINI: Thank you.
12 MS. WEBBER: You're welcome. All right.
13 I'm going to turn it over to Don now.
14 MR. ALGAMA: Thank you. Hopefully I can 15 change. Oh, there we go. Can everyone see the 16 slides?
17 MS. WEBBER: Yes.
18 MR. ALGAMA: Thank you. Howdy. My name 19 is Donald Algama and I'm with Drew Barto. Today we're 20 here to discuss Volume 5 as Kim as already provided.
21 It is important to note that this is a 22 plan. And as we learn more during the process, 23 especially implementation and gathering information 24 from the DOE and vendors, we will update the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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26 implementation part of the plan as we move forward.
1 Sorry. It starts changing. Oh, there we 2
go. Okay. This is an acknowledgment to all the great 3
help we received from both the program officers from 4
NMSS, NRR and research and also David Luxat from 5
Sandia and Will Wieselquist from Oak Ridge. So thanks 6
for all the help in doing this.
7 You've already seen this part so I'll 8
skip over this. This is just a summary of the IAPs to 9
date. And with this, we start.
10 The goal is to apply and understand the 11 performance of existing NRC tools to support fuel 12 cycle evaluations. And the intention is that we will 13 gain experience in all fuel cycles and at the same 14 time demonstrate computer code readiness.
15 As a plan, it is intended to be updated as 16 we learn more from DOE and the industry for both the 17 designs and what they may be expecting from their 18 normal fuel cycle approach.
19 This plan will take on a delta approach 20 using the existing LWR fuel cycle as a reference.
21 Basically, an incremental approach comparing the 22 candidate and non-LWR design against existing fuel 23 cycle capabilities.
24 As we are taking an LWR approach, in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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27 practice this means core knitting with internal 1
partners when scenarios demonstrate the need such as 2
those in Volume 3 and Volume 4 and our NMSS teams 3
concerned about release, dose, materials, et cetera.
4 Volume 3, the impacts using this work will 5
be made public. This plan leverages LWR experience to 6
the extent possible. Thus, the following few slides 7
will provide an idea of how these codes are used in 8
the existing framework and existing staff experience.
9 The red box highlights areas in the LWF 10 fuel cycle as a potential use in this work. The 11 following two slides will provide further examples.
12 This slide provides an overview of the 13 transportation of storage space as of today. The 14 slides start from fundamental nuclear data, processing 15 the application to scale and then possible follow-on 16 work.
17 In this area, scale is currently being to 18 the context of criticality and shielding for spent 19 fuel package designs and for spent fuel dry storage 20 systems, shield analysis to support radioactive 21 material process and package designs and for dry 22 storage systems including the waste consolidation 23 storage and Holtec HI-STORE Consolidated Interim 24 Storage Facility applications.
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28 It's also been used in transport, 1
criticality analysis for packages of UA6, U02 powder 2
and pellets, commercial and research, fresh and spent 3
fuel assemblies, et cetera.
4 MEMBER MARCH-LEUBA: Don?
5 MR. ALGAMA: Yes?
6 MEMBER MARCH-LEUBA: This is Jose.
7 MR. ALGAMA: Hi, Jose.
8 MEMBER MARCH-LEUBA: Yes. Have you 9
thought about the uncertainty of core second 10 generation? For a long time core second generation 11 was an art. It has now become more of a science but 12 that's because of all the experience we have with 13 configuration with fuel rods and light water. And we 14 have resolved all the problems.
15 But when you are going to these unusual 16 configurations like a molten core or even a little bit 17 of the pebble reactors. So have you given 18 consideration to uncertainty of cross-sections?
19 MR. ALGAMA: Yes. That will be considered 20 in the implementation phase in part of the 10 reports.
21 MEMBER MARCH-LEUBA: And is there going to 22 be sufficient data to benchmark criticality?
23 MR. ALGAMA: Yes and no.
24 MEMBER MARCH-LEUBA: Okay.
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29 MR. ALGAMA: So as of right now for the 1
HALEU space, we are developing approaches to mitigate 2
the lack of benchmark data or appropriate benchmark 3
data, but we'll be evaluating those as we go through 4
the implementation phase.
5 Will Wieselquist can answer more if he 6
can, but we'll be evaluating it. But we haven't 7
really got there yet.
8 MEMBER MARCH-LEUBA: Okay, yes. You need 9
to give it some thought because if there is need for 10 experimental data for a
particularly unusual 11 configuration for which we don't have any experience 12 that would be really bad because we --
13 (Simultaneous speaking.)
14 MR. ALGAMA: Yes. Understood.
15 MR. BARTO: So this is Drew Barto. I 16 don't think Will is on the line. But I can try to 17 answer for him. You know, that is a very good point.
18 And that's a big part of what we'll be looking at in 19 terms of gaps. You know, really moving forward we've 20 used these tools for a number of years, you know, 21 mostly for LWR type of analyses.
22 But we really have been able to evaluate 23 some of the materials and configurations that are 24 going to be used in the advance reactor fuel cycle.
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30 So we've been able to -- as far as the codes 1
themselves, they have the capability of modeling these 2
things, like, you're right. None of that means 3
anything if you can't validate it.
4 And so that's a very important part of 5
what we'll be looking at. You know, what experiments 6
are available? You know, to what extent can you use 7
experiments?
8 You might now think it looks like your 9
system, but neutronically they are similar so there's 10 lots of use of say, sensitivity and uncertainty 11 analyses, methodologies to compare critical systems.
12 So, you're right, that is a very important 13 part of this.
14 MR. PETTI: So are you guys hooked into 15 the criticality benchmark, IAEA activity where they 16 have housed tremendous amounts of data on criticality 17 and other similar experiments across the reactor 18 spectrum so there's been tons of gas reactor stuff 19 that I'm aware of, fast reactor stuff that you guys 20 could, you know, check tools against?
21 MR. ALGAMA: I will look into it. I'm not 22 aware of this off the top of my head.
23 MR. PETTI: It's a huge -- I mean, it was 24 a, I don't know, three or four person effort probably 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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31 in the U.S. alone, and it's international in its scope 1
so.
2 MR. ALGAMA: Is it different than to the 3
OECD benchmark?
4 MR. PETTI: No, no, no. I'm sorry. OECD 5
is what I meant, not IE --
6 MR. ALGAMA: Oh, yes, we're aware of that, 7
yes.
8 MR. PETTI: Yes, yes. There's a lot in 9
there so.
10 MR. ALGAMA: Yes, sir.
11 MR. PETTI: Yes.
12 MR. ALGAMA: And we used that in part of 13 our valid suite, too, for validating scale or setting 14 scale's performance.
15 MEMBER REMPE: Don?
16 MR. ALGAMA: Yes, ma'am.
17 MEMBER REMPE: This is Joy. I had a 18 question or comment. I was looking through the 19 report, and I'm not sure how you would address it, but 20 I think a paragraph is worthwhile to add to the report 21 about these reactors that are supposed to be 22 fabricated in a different facility and the core loaded 23 and then transported and installed at a site and then 24 removed from the site and taken somewhere for whatever 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 they do to unload the fuel.
1 Because I assume it would be covered in 2
this Volume 5 activity, but it's not really discussed 3
or I missed it if it was discussed in the report and 4
what you plan to do on it. And I'm not sure what you 5
would do, but perhaps it ought to be acknowledged that 6
this something that may have to be considered.
7 MR. ALGAMA: Understood.
8 MEMBER REMPE: But what are your thoughts 9
about what you would do with something like that?
10 MR. ALGAMA: Going through the fuel cycle, 11 I think the intention was the -- I think the tables --
12 we provided the flowchart of analysis within.
13 MEMBER REMPE: Right. And I --
14 MR. ALGAMA: That would be where we 15 discussed those kinds of activities. So we start --
16 MEMBER REMPE: So I looked for that, and 17 I did not -- again the way the sodium fast reactor 18 because one of the ones they're talking about, I did 19 not see it there or in any of the others where it just 20 called out and said we need to think about this type 21 of structure where you would actually have -- they 22 talk about loading the core at the site. They don't 23 talk about loading it offsite and transporting it to 24 the site, right? I did not see that in one of those 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 flow diagrams.
1 MR. ALGAMA: I understand. So that was 2
the difference between the HPR and the SFR cores, 3
where the SFRs had a, like, a regular LWR approach 4
where their centers would be manufactured and then 5
shipped out to the site for loading. And then the HPR 6
where we anticipate that the whole reactor core will 7
be fabricated in the fabrication site and then shipped 8
out.
9 We did try to put some text in the report 10 about the two different approaches, but we can add 11 more to be --
12 (Simultaneous speaking.)
13 MEMBER REMPE: Maybe I missed it. But, 14 again, I think that that is something that may -- I 15 mean, do our existing tools cover something like that?
16 MR. ALGAMA: Existing tools cover -- I'm 17 not sure. Forgive me. Could your rephrase the 18 question?
19 MEMBER REMPE: Well, do we think about 20 transporting -- I mean, can you use scale or something 21 to deal with a criticality event when you have a 22 loaded core being transported somewhere and installed 23 on the site?
24 MR. ALGAMA: Yes.
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34 MEMBER REMPE: I mean, because we have the 1
tools and capabilities for doing that we just haven't 2
ever applied them for such a situation?
3 MR. ALGAMA: Correct. Yes, we can apply 4
the tools. But like Jose was saying, we have to be 5
careful on what the results mean, developing an 6
appropriate validation basis and uncertainty analysis 7
to go with it. But yes, the short answer is yes.
8 MEMBER REMPE: Okay. So I just think that 9
we need to discuss that a bit more in the report to 10 acknowledge that we're thinking about it, but, you 11 know, it's something that will be addressed or 12 something. You know, I guess I did not see that 13 enough when I was looking in the text but maybe I 14 missed it.
15 MR. ALGAMA No. We can add more. Thank 16 you.
17 MR. PETTI: Okay. This is a case again 18 the assumption on the heat pipe reactor, I understand 19 where it came from. But there's another heat pack 20 reactor potentially, at least a microreactor that it's 21 different enough that it may cause you to rethink a 22 little bit how the different pieces fit together.
23 And that's what I kept struggling with is 24 in general you have to make a number of assumptions, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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35 right, to kind of weigh this out. What if you're 1
assumptions are wrong and how would that impact, you 2
know, the approach? It would just seem like it would 3
be worth a little bit of thinking about that. I don't 4
think it will change the fact that the tools, you 5
know, can do the job. It's just, you know, your view 6
of the future may not be exactly what the future is.
7 MR. ALGAMA: Understood.
8 MR. PETTI: Right. So, I mean, it might 9
be worth just a paragraph or even a footnote of that 10 that, you know, even though this is what we've said, 11 we think, you know, more broadly that the tools can 12 handle, you know, some sort of evolution away from 13 these assumptions so.
14 MR. ALGAMA: Yes, sir.
15 MR. BARTO: Hey, this is Drew. And I'll 16 just add to that. I think you're right, it could 17 benefit from a little more discussion. And I think as 18 far as neutronics tools for criticality and shielding 19 that it's not going to be that much of a challenge to 20 model, you know, whatever comes forward in terms of 21 heat pipe reactors or other transportable reactors.
22 The challenge with those is really going 23 to be in the structural and thermal analysis showing 24 that they can survive the 10 CFR Part 71 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 transportation accidents, which I'm sure you're aware 1
are much more challenging for a stationary system.
2 So it's going to be showing that the 3
system can withstand those accidents and then 4
translating that into a configuration that your re-5 tracks tools can model. And is that configuration 6
appropriate? And that's really going to boil down to, 7
I think, the nuts and bolts of an actual technical 8
review. But it should not be a challenge for the 9
scale or the other tools to model such configurations.
10 MR. PETTI: Right. Thanks.
11 MS. WEBBER: But the one thing I want to 12 note. I do agree that it's worth adding, you know, 13 some information about that configuration, you know, 14 with the fuel loaded into the reactor and then the 15 whole reactor with the fuel shipped to wherever it's 16 going. So I think that's something that we can do.
17 MR. ALGAMA: It's more of a story of what 18 we anticipate and how we would accommodate changes.
19 MS. WEBBER: Well, the nuances of that 20 particular type of reactor design, microreactors.
21 MR. ALGAMA: Okay. I'm going to move to 22 the next slide. Is that okay? I take that as a yes.
23 Just so that I capture the basis of Volume 24 3 approach from our analysis as you've seen before.
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37 As before, fundamental data is processed and applied 1
by SCALE and passed as input a severe -- as input of 2
a severe accident and source term code MELCOR and 3
offside analysis code MACCS.
4 The following slides are some examples of 5
starting fuel cycle experience applying the scale of 6
MELCORs to non-reactor facilities in transport and 7
storage areas.
8 The codes have been applied in the L3 PRA 9
project. And here at 2161 is the spent fuel core 10 study at NUREG 7108 and 7109, which is the developing 11 estimates on isotopic depletion bias and uncertainty 12 and criticality uncertainty.
13 This is a recent application of scale in 14 MELCOR to a non-power facility. This analysis looks 15 at a range of scenarios at the Barnwell Nuclear Fuel 16 Plant and the effectiveness of various plans of 17 defense within the reprocessing facility.
18 Five of the classes of accidents in the 19 FSA were evaluated with the scale MELCOR package. And 20 we captured material degradation, building leakage, 21 aerosol physics for deposition, agglomeration, et 22 cetera. And we also looked at leak path factor 23 considerations, impacts of filters, ventilation 24 systems, instructs as a result of fires.
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38 MEMBER PETTI: I had a question back on 1
the burner credit. You know, some of these burnups 2
significantly beyond what we think of in the light 3
water reactor context.
4 MR. ALGAMA: Yes, sir.
5 MEMBER PETTI: Do you guys have any idea 6
how good the SCALE code suite will do? Because, you 7
know, you're going to be fissioning a lot more 8
plutonium as you get those really high burnups and the 9
uncertainties of the fissioning of the higher 10 actinides?
11 MR. ALGAMA: Yes. So we're actually 12 pursuing research as part of ATF/HBU to see if we can 13 develop methodology that would extend or depletion and 14 uncertainty analysis along with that.
15 We would eventually need validation data 16 to see just how good we are, but we have an approach 17 in mind.
18 MEMBER PETTI: So there is data, very 19 recent data, for gas reactors. And I think there's 20 probably similar data for a fast reactor fuel as well.
21 So it's just a matter of getting access to it.
22 MR. ALGAMA: Yes, sir. You wouldn't by 23 chance have the reference for that do you?
24 MEMBER PETTI: Well, the HER program has 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 published the burnup comparisons with actual 1
destructive burnup and measurements of season fission 2
product ratios correlated to burnup. So that's out 3
there in the public literature. And the fast reactor 4
stuff is a little bit older because we haven't had a 5
fast reactor in the U.S. But I'm sure there's data 6
from EBI, too --
7 MR. ALGAMA: Yes.
8 MEMBER PETTI: -- that would be useful so.
9 MR. ALGAMA: I see.
10 MEMBER PETTI: Yes.
11 MR. ALGAMA: Thank you. Let's skip over 12 this one. So this slide is a copy of Table 1-1. The 13 intention is to provide a high level of understanding 14 of what differentiates non-LWRS and LWRS right now.
15 Some notable features are that the designs 16 are based on uranium and share front end UA6 17 enrichment needs that are common and some fabrication 18 needs that are common.
19 Fuel forms range from oxides and metals to 20 uranium dissolved in molten salts. The neutron 21 spectrum can be firm all the way to fast. Burnups, as 22 you mentioned, Dr. Petti, can be very large compared 23 to LWRs and numbers that potentially include onsite 24 fuel processing. So all these things will have to be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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40 evaluated.
1 As mentioned the objectives in this plan 2
and its resulting reports ultimately demonstrate 3
computer code readiness. To achieve this, we will 4
have to look at developing scenarios and identify 5
potential hazards to assess the codes against.
6 We intend to look at available NRC, DOE 7
and design information as they come up to help 8
understand the potential on non-LWR fuel cycle. And 9
thus this plan will evolve as we implement as well as 10 historical information.
11 MEMBER REMPE: Don?
12 MR. ALGAMA: Yes.
13 MEMBER REMPE: I didn't meant to interrupt 14 you. Go ahead and finish. But I have a question when 15 you finish this slide.
16 MR.
ALGAMA:
- Yes, ma'am.
Hazard 17 evaluation, there are documents that can be used to 18 develop scenarios to test core performance in 19 criticality safety, our inventory characterization 20 indicate heat estimation, radiation shielding and RN, 21 radionuclide and other hazard evaluations.
22 Further analysis needs -- consequence 23 analysis areas will be raised to the appropriate team 24 at NSNRI within Volume 3 and 4 as they occur.
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41 We will use NUREG 6410 to drive our 1
scenario selection for fuel cycle facilities. And in 2
particular, it includes a process hazard analysis 3
approach, which is a technique to identify and 4
understand scenarios that merit further analysis.
5 This handbook, 6410, covers criticality 6
events, release of materials, in-facility transport 7
depletion processes, leak path factors. And Table 2 8
of that provides a range of scenarios that could be 9
considered for existing facilities.
10 In 1520, which compliments 6410, the 11 purpose of the SRP is to ensure quality and uniformity 12 of reviews, which also provides further insights on 13 how we should assess our codes.
14 In 2015, the move from facilities to 15 transport. And this NUREG focuses on COC for dry 16 storage systems and ISFSIs and monitored retrievable 17 storage installations.
18 In 2016, we moved towards transportation, 19 which covers fueling criticality, et cetera, and 20 provides a -- Table 1-2 of this report provides an 21 example of scenarios to demonstrate some criticality.
22 And Attachment 2A provides staff expectations of 23 computer codes.
24 Moving along, there are complementary DOE 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 documents that we could leverage. One as an example 1
that may be useful to develop hazards is listed. The 2
other documents such as DOE Standard 1027 has an 3
evaluation techniques, DOE Standard 2007, which covers 4
SERs for non-power facilities, et cetera. These will 5
be all reviewed in the implementation phase.
6 So an example scenario may be an accident 7
at a fuel fabrication facility. An accident occurs 8
where -- I hypothesize, where the UA6 cylinder is 9
damaged while it is in the process of being evacuated.
10 Staff may be interested in investigating possible UA6 11 release, chemical reactions from the damaged canister 12 and into the facility environment.
13 Joy, I'm going to move to the next slide 14 so you had a question?
15 MEMBER REMPE: Yes. First of all, earlier 16 I meant to tell you I really like Slide 5 and Slide 17
- 10. I thought those were nice slide summaries of how 18 codes were used for those regulatory activities and 19 where there were gaps.
20 But when I was looking in your report and 21 thinking about how you're going to develop scenarios, 22 I think it might behoove NRC -- I'm not as familiar 23 with this DOE handbook. But it might behoove NRC 24 staff to think about a more in-depth review of prior 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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43 experience that's more recent.
1 The Tokaimura accident happened in 1999 2
but 6410 was a lot older as I recall. You mentioned 3
you've got a lot of experience, the Agency does, with 4
non-LWRs and you go back and mention this being an L 5
report. But it's a very high level summary report 6
that rarely go into depth of things that have happened 7
with gas reactors like Fort St. Vrain as well as 8
Fermi.
9 And there are a lot of times where lack of 10 administrative controls have led to fuel melting and 11 severe situations like what happened at Tokaimura.
12 And I am wondering if maybe some more in-depth review 13 is needed unless there's something in this DOE 14 Handbook that will give you some really good ideas 15 about scenario selection. What are your thoughts 16 about that?
17 MR. ALGAMA: No, no. I one hundred 18 percent agree. That was the intention also was to 19 look at historical data to guide us in what would be 20
-- hazards of interest to apply our codes and see how 21 they perform.
22 MEMBER REMPE: Yes. Because I do think 23 there's some very good lessons in history. But I just 24 haven't seen enough discussion of that. And so it 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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44 might behoove you to go a little more in-depth. Bad 1
things have happened when people do things without 2
enough review and don't have enough administrative 3
controls. And I'll stop there.
4 MR. ALGAMA: Yes, ma'am.
5 MS. WEBBER: So, Joy, just to make sure I 6
understand your comment. So are you suggesting that 7
in the report that there's maybe a little bit more 8
about scenarios that need to be evaluated in the 9
context of the scope of the report?
10 MEMBER REMPE: I think the report is fine.
11 But I think maybe research might want to think about 12
-- again it depends on how the future plays out. But 13 if we're going to try and do this for non-LWRs, I 14 think a more detailed review of what's happened in the 15 past would behoove us.
16 MR. ALGAMA: Could I just state one -- I'm 17 sorry.
18 MEMBER REMPE: Yes. And then, again, when 19 you don't have the details of these new facilities 20 because they're just conceptual ideas, it's hard to do 21 that. But I think those things -- you know, again, I 22 recently was involved in a project where we looked 23 more in-depth of what happened at Fermi 1 and Fort St.
24 Vrain with its startup.
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45 It's just when there's not enough 1
administrative controls, there's not enough review, 2
things have happened. And Tokaimura is an example 3
where, again, people applied something, a process they 4
had used for a lot of times to something a bit 5
different. And people didn't, you know, have enough 6
oversight and review of the situation before things 7
occurred.
8 And so, again, I was interested in your 9
report. And you mentioned, oh, you've got this 10 Brookhaven report. And there's barely a paragraph 11 about each reactor.
12 And I think somebody needs -- I'm sure 13 there's people around, and there's a lot of history 14 around. And I just think it might be a good thing for 15 research to do if this whole non-LWR thing comes to 16 fruition.
17 MR. ALGAMA: Would that be something we 18 would consider an implementation phase? That was the 19 idea at least.
20 MEMBER REMPE: Yes. I think, I mean, you 21 might acknowledge that clearly a more in-depth review 22 would be performed because of situations in the past.
23 But I just think that a more detailed review would be 24 good.
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46 And how you want to address that, again, 1
I wouldn't go spend money on it today unless we know 2
for sure somebody is going to do this, but I think a 3
more detailed is needed at some part. And it's up to 4
you guys how you take that. It's just one member's 5
comment if you want to try and do something that way.
6 MR. ALGAMA: Understood.
7 MS. WEBBER: To me it sounds like really 8
a, you know, broader operating experience review of 9
all the technologies.
10 MEMBER REMPE: Yes.
11 MS. WEBBER: Okay. Thanks. I'm not sure 12 it's really in the scope of this report. But where 13 it's relevant, you know, we could, you know, add some 14 additional text.
15 MR. ALGAMA: So once we are done with 16 scenario selection, we move on to the scope of the 17 analysis. With areas such as mining, milling, long-18 term storage and disposal consequences, radiation 19 protection, chemical toxicity would be counted 20 elsewhere.
21 CHAIR BLEY: I'm sorry. But my brain just 22 caught up with --
23 MR. ALGAMA: Yes, sir. Do you want me to 24 go back a slide?
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47 CHAIR BLEY: No. This is for Kim and our 1
past discussion. If we're looking at scenarios and 2
the ability to identify them is crucial and if we 3
don't look carefully at the history when missing a 4
source of information to make that a more complete 5
assessment, I don't see why it doesn't fit here, Kim.
6 MS. WEBBER: Yes, I guess. So in the 7
context of the front end and the back end of the fuel 8
cycle, you know, I think, you know, there's obvious 9
relevance to this scope.
10 But I think what Joy may have been 11 advocating, and correct me if I'm wrong, is something 12 more broad about, you know, she mentioned admin 13 controls and startup of the reactor. And so there's 14 broader operating experience related to the operations 15 of these reactors.
16 And so I think that the, you know, really 17 what's relevant to the fuel cycle are the operating 18 experience relative to the front end and back end of 19 the fuel cycle. I think that's what I meant.
20 CHAIR BLEY: Okay.
21 MS. WEBBER: But thanks for the comment.
22 I appreciate that.
23 MR. ALGAMA: All right. As with Volume 3, 24 we expect to reasonably apply comprehensive 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 methodological approach from scenario definition, 1
identification of safety related items, identification 2
of dominant phenomena to support that through the V&V 3
and documentation.
4 We also intend on using the designs 5
developed in Volume 3 to support fuel cycle analysis 6
in Volume 5.
7 Continuing an example, it continues from 8
the previously mentioned. Staff may want to know how 9
the UA6 can be transferred in the damaged canister, 10 how much HF is produced and where is the uranium 11 deposited within the facility, specifically the HVAC 12 to understand criticality implications, deposit 13 materials, et cetera. We would deploy a combination 14 of SCALE and MELCOR to try and evaluate that scenario.
15 Here, we move on to the 10 anticipated 16 reports. Obviously, this would all be contingent on 17 what we learned. We can adapt. We are flexible. As 18 we learn more from the DOE and its partners, we can 19 change how we prioritize the work in both 1, 3 and 5.
20 The term reports are broken down into five 21 reports looking at non-LWR, specific fuel cycles and 22 five reports that cover common fuel cycle activities.
23 The reason for this is to take advantage 24 of commonalities. If you look at the HTGR and FHR 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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49 fuel cycles, we can see that Reports 3, 7 and 10 are 1
common. So once developed for one, it will be 2
applicable to the FHR, for example.
3 MEMBER PETTI: So let me just -- if you go 4
back. This is a common flaw throughout the whole 5
report, a nomenclature problem, on Number 7 here, 6
TRISO fuel kernel. The kernel as a nomenclature is 7
the fissile part of the particle. But I'm sure you 8
would read about the particle fabrication as well.
9 MR. ALGAMA: Yes.
10 MEMBER PETTI: So do you think you want to 11 say kernel/particle or kernel and particle fabrication 12 and just go through the whole report. And most of the 13 time I think you mean particle. But there are a 14 couple of times where I think you meant both, the 15 fissile kernel and then the coated particle, just to 16 use nomenclature that's more traditional.
17 MR. ALGAMA: Yes, sir.
18 MEMBER PETTI: Similarly, this is one of 19 the assumptions that struck me was that you assumed 20 that the fuel element here, you have it as a pebble, 21 would be a different facility from where the particles 22 are made. That has never, ever happened in the world.
23 All of the Germans, the Chinese, the Japanese, the 24 Americans all -- it's all in one facility.
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50 There can be different material balance 1
areas for sure to deal with accountability and the 2
like, but they would not probably be large scale 3
shipment of coated particles from one facility to 4
another because they are actually fairly fragile in 5
that state. And so it's always done in one facility.
6 MR. ALGAMA: Yes, sir.
7 MEMBER PETTI: So I would clean that up 8
just so, you know, people wouldn't say, oh, they don't 9
really know what's going on.
10 MEMBER KIRCHNER: Yes. Dave, this is 11 Walt. I agree, yes. The nomenclature on seven should 12 be more inclusive. And, yes, 10 as a standalone, then 13 it begs the question what about compacts, which is the 14 alternate means of taking the particle fuel and 15 putting it into a serviceable form that can be loaded 16 into a reactor.
17 MR. ALGAMA: Right.
18 MEMBER KIRCHNER: So, yes, I think these 19 could be combined.
20 MEMBER PETTI: And then, you know, Kim is 21 talking about shipping TRISO fuel. And it's compact.
22 And that's a project that's underway right now. And 23 so this is a case where you guys are trying to see the 24 future, and, you know, it doesn't align with where we 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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51 are today. So you could just say compact or pebble 1
fabrication and --
2 (Simultaneous speaking.)
3 MEMBER KIRCHNER: Yes. I would combine 4
them. When I bought fuel from GA, it was shipped to 5
us in the form of compact. So it wasn't loose pebble 6
particles.
7 MEMBER PETTI: Particles, right, right, 8
so.
9 MR. ALGAMA: We didn't actually consider 10 transport of TRISOs to a pebble facility. Will is on 11 the line right now maybe he can add to this. But we 12 did try to make a differentiation between pebble and 13 fuel compact scenarios for the fuel cycle. Will, can 14 you chime in a little bit? But we can make updates to 15 the report to make it clear.
16 MEMBER PETTI: Yes. It would be 17 interesting to know why you thought there was a 18 difference, at least at the level that you guys are at 19 20 MR. ALGAMA: Mm-hmm.
21 MEMBER PETTI: -- they look really 22 similar. If you would have recycled the fuel in type 23 of a cover uranium, things can get a little bit 24 different. But they go through all the same steps.
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52 It's just the geometry is instead of pressing a 1
cylinder you're pressing a sphere.
2 MR. ALGAMA: Okay. And we referenced 3
compacts, but we didn't look into it because at the 4
time of this report we didn't have a driver for it.
5 But that's something we can look at again.
6 MEMBER PETTI: Right. And now this one 7
microreactor project the basis is TRISO and compacts.
8 MR. ALGAMA: Yes.
9 MEMBER PETTI: And then again, that's a 10 thermal system. That's another thing that when you 11 mentioned heat pipe reactor, you basically locked 12 yourself into fast, a fast system, but they are 13 thermal systems as well.
14 MR. ALGAMA: Understood.
15 MS. WEBBER: Thanks, Dave and Walt. I 16 appreciate those insights.
17 MR. ALGAMA: So this leg, we begin our 18 strategy. As mentioned, the LWR fuel cycle we use as 19 a reference to understand the anticipated non-LWR fuel 20 cycle. To make the task more tractable, we broke them 21 down into six major steps and several stump steps.
22 These are labeled with the first step of 23 the stage and a number for the substep.
24 So for fabrication we can break down the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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53 two steps, identify the F1 and F2. This work will not 1
right now look at scenarios of interest in the T3 and 2
S1 steps due to lack of understanding of where the DOE 3
industry plans to go. That's probably way, way too 4
far in the future for us. We will revise as we learn 5
more.
6 The FHR class, the fuel cycle analysis, 7
will be driven by the Berkeley Mark-1 FHR design as we 8
had in Volume 3. The basic design uses TRISO 9
particles up to 20 weight percent.
10 This directed design loads pebble from the 11 bottom and are removed from the top. There are 12 hundreds of thousands of pebbles that are expected to 13 be used with thousands of TRISO particles each.
14 Rather than helium they will use a molten 15 salt like FLiBe as the coolant. But the fuel cycle 16 analysis stage, I expect it to be identical for what 17 do for HTGRs but with some additional features such as 18 moats for fission particle inventory migration within 19 the coolant and then compared to HTGRs and tritium 20 generation, transport and retention phenomena in both 21 the FLiBe and the graphite.
22 Steps E1 and E2 will be completed in 23 earlier reports as we described for commonalities. In 24 E1, we will look at fresh fuel, how they will be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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54 staged and the expectation is looking at criticality 1
type accidents here from fuel handling operations.
2 Step E2 is covered in Volume 3 where 3
interactive data such as anticipated discharge relapse 4
will be generated. This work may also consider 5
radionuclide hazards during different fuel cycle 6
operations and hazards with respect to fuel handling 7
as I mentioned earlier.
8 In Step U3, it is not expected because we 9
don't expect central fuel shuffle operations.
10 In Step 4, we expect onsite storage of 11 spent fuel pebbles will be reviewed with respect to 12 criticality, fuel and decay heat and other accidents.
13 For the HPR fuel cycle, it will be driven 14 by a modified version of INL Design A, which comes 15 from Volume 3. The basic design is the SFR and HPR 16 are essentially the same in the front end of the fuel 17 cycle, with the exception of how the fuel is actually 18 manufactured.
19 Traditional SFRs have assemblies while 20 HPRs are expected to be manufactured as an entire core 21 but a bit smaller than an SFR core.
22 The fuel will be modified to be metallic.
23 The INL design and discharge burnups increase around 24 10 gigawatt day MTU (phonetic).
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55 The design is hexagonal with a sodium bond 1
that thermally connects -- with a sodium bond to 2
connect the fuel and the coolant.
3 In Steps E1 through F1, it will be done 4
earlier. The work will start at the F2 stage, 5
fabrication of the HPR core to reach transport to the 6
utilization stage.
7 The F2 stage included the step due to the 8
unique processes we anticipate when you're looking at 9
developing a whole new core to transport.
10 The new stage of the core, the fresh core 11 will be reviewed with respect to criticality concerns, 12 staging areas, et cetera.
13 Stage U2 will make use of developments in 14 Volume 3 and again also vary and are adapted for use 15 in metallic uranium.
16 In the U4 stage, we will look at the full 17 range of criticality shielding decay heat and hazard 18 analysis.
19 The SFR fuel cycle reference reactor is 20 under consideration still. Two possibilities stand 21 out as the MET-1,000 benchmark design or the VTR.
22 More information will be reviewed as we go into the 23 implementation phase for this phase of the report.
24 Basic information is that this design can 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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56 come with a wide range of fuel fonts from oxides, 1
carbides, nitrides and metals. The metallic form will 2
likely be a driver for this work. Enrichments up to 3
20% can be expected.
4 As before, Steps E1 through T2 will be 5
covered in other reports. At U1 stage, we will look 6
at criticality concerns mainly we anticipate for the 7
fresh fuel assemblies. At the U2 stage, we will 8
leverage the work that will be performed under Volume 9
3.
10 Unlike the HPR, we do anticipate the U3 11 stage to understand accident scenarios with spent fuel 12 shuffling operations.
13 With U4, we expect to review the full 14 gamut of technical areas as mentioned before with both 15 scale and melt core.
16 MEMBER PETTI: So just so I understand, U3 17 you mean shuffling in core like we do in light water 18 reactors?
19 MR. ALGAMA: Yes, sir.
20 MEMBER PETTI: Okay, okay.
21 MEMBER REMPE: And, Don, if you'll go back 22 a slide? Okay. So this is why, and I think Dave 23 captured it correctly by saying this is a bit 24 different than the folks that are thinking about 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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57 putting the core in the vessel or some container and 1
installing the whole reactor vessel at the site.
2 And so perhaps this is one type of a heat 3
pipe reactor, but there are other types where you have 4
a fully loaded core that you move to the site. And 5
that's not reflected in this diagram on your report, 6
right?
7 MR. ALGAMA: Yes, ma'am. That's correct.
8 When we started this work, we really looked at the 9
designs that were being evaluated in Volume 3, and we 10 used that to drive this report because we thought that 11 was a good representation of what might come forth in 12 the near future.
13 MEMBER PETTI: This is why I think a 14 footnote to recognize that there are other options.
15 MR. ALGAMA: Yes, sir.
16 (Simultaneous speaking.)
17 MEMBER PETTI: -- if you can change the 18 whole, you know, strategy of the report. But it's 19 just that, you know, you could say, yes, we're aware 20 of that other thing over there so.
21 MEMBER REMPE: So, yes, I think especially 22 because I think Amy Cubbage mentioned this at a 23 stakeholder meeting last month maybe, actually October 24 or November, I forgot now which month. But she talked 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 about that this might be a policy, challenge some 1
policy issues. But it's something that the Agency 2
needs to observe and note that they are aware of this, 3
and they are starting to think about it.
4 MR. ALGAMA: Yes, ma'am. I'm going to go 5
to the issues here. For this analysis, we will be 6
using PBR 400 as in Volume 3.
7 This information is from NGNP, in other 8
words that we know there are two types of HTGRs we can 9
look at though in the form of pebble bed and prismatic 10 type. The main difference between the two is expected 11 to be with the fuel utilization stage, however, where 12 the pebble bed design is not expected to have a U3 13 stage for fuel shuffling, used fuel handling 14 inspection, et cetera.
15 For the PBR 400 though we expect what 16 will drive this work from Volume 3, we expect about 17 400,000 pebbles each with tens of thousands of TRISO 18 kernels within the reactor core, and helium is used as 19 the coolant.
20 As far as the approach, this will look 21 just like the FHR section that we just discussed. For 22 MSRs, currently we're looking at the MSRE as the 23 driver for this fuel cycle report.
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59 along with models already developed within Volume 3 1
but not much involving the fuel cycle. This will have 2
to be more of a research activity in which fuel salts 3
can be transported to the site and diluted with salt 4
available onsite before using in the reactor for 5
example. More work needs to be done from a fuel cycle 6
perspective.
7 As before, E1 and F1 are addressed 8
elsewhere because there will be a UA6 initial phase.
9 F1 fabrication step is looking at fabricating UA6 into 10 uranium dissolved in salt in which fuel salt 11 manufactured at F1 step is expected to be transported 12 to the site where it would combine with fuel salt at 13 the site and hydraulically transferred to the reactor 14 circuit.
15 This stage will focus on actions that 16 we're looking at criticality, chemistry use, et cetera 17 there.
18 In the U1 step, we will look at 19 criticality, shielding and issues and operations such 20 as blending, handling, et cetera.
21 And in the U2 stage, power production, 22 unlike chemical processes, will be covered in Volume 23
- 3. But refueling and processing capabilities are 24 expected to be needed to remove salt and extract 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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60 fission gas during operations. So that might be 1
covered in this report and in Volume 3 as appropriate.
2 In the U4 stage, effort will be spent at 3
criticality issue being regular transport and other 4
chemical processes of interest that we identify.
5 This work has all other areas that we 6
intend to make use of. From the front end UA6 works 7
for the ATF inherent work. There are commonalities, 8
and we will leverage those as much as we can.
9 So Volume 3 we will leverage the reference 10 designs developed there and companion work to 11 understand nuclear data -- and companion work that is 12 being utilized to understand nuclear data performance.
13 This is useful as this not only helps 14 define the fuel cycle for what we're going through but 15 the radio fuel characteristics that drive the back 16 end.
17 In the implementation phase, we also are 18 intending on expanding collaboration with the DWD re-19 programs that are in this area upon the start of the 20 work.
21 We are aware the DOE expects a certain 22 amount of time looking at various fuel cycles, the 23 efficacy of the fuel cycles and a number of reactor 24 designs.
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61 In conclusion, being we had a reasonable 1
approach, a reasonable strategy in the reference to 2
delta strategy benchmarked against the LWR fuel cycle, 3
we believe that the development assessment work being 4
performed under Volume 3 will help cover the 5
development needs in Volume 5 so we don't expect new 6
phenomena that aren't already captured in our codes.
7 What we're mainly focusing on is 8
understanding how to revalidate our codes and what 9
does that mean when we have more or less or in between 10 months of validation data, whether we can mitigate the 11 lack of data by using new methods and where we will 12 just have to have new data available.
13 We believe that sufficient experience in 14 the application of SCALE and MELCOR to non-reactors 15 exists to start the process. But this experience will 16 be developed and refined as we get more experience and 17 implementation and also from DOE industry.
18 We will leverage other NRC programs to the 19 extent possible, including Volumes 3 and 4 as the 20 scenario dictates. That's all I have today. Thank 21 you.
22 CHAIR BLEY: Thanks. Kim, do you have 23 anything more?
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62 Dennis, not specifically.
1 CHAIR BLEY: Members, if you have any 2
questions, bring them up now please. After public 3
comment, I'm going to go around and have everybody 4
discuss a couple of things. But is there anybody on 5
the committee who wants to ask any more questions at 6
this point?
7 MEMBER PETTI: So, yes. I had one. I'm 8
still struggling with after fabrication -- there is 9
only one fabricator in the country today that can 10 handle HALEU material that has a license from the NRC.
11 So this is, again, one of these assumption 12 things. They already have a license. So they can do 13 a lot of stuff, and it may not actually require, you 14 know, an NRC review.
15 MR. ALGAMA: I see.
16 MEMBER PETTI: Because they have all of 17 the, you know, safety paperwork in place.
18 It's probably worth talking about 19 somewhere just, you know, what would have to happen to 20 stand-up, you know, a fabrication plant that can 21 handle HALEU. It's a lot different than LEU, you 22 know, LWR fuel, whether that be modifying, you know, 23 an LWR fuel vendor to allow them to handle HALEU or 24 not so if someone wants to get into the game, you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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63 know, brand new.
1 MR. ALGAMA: Would this be an extension to 2
this work? The whole idea was to try and show core 3
readiness with this --
4 MEMBER PETTI: Yes. To me, it's just a 5
footnote so you guys recognize that there are 6
different options. One is a current LWR fuel vendor 7
wants to make these advance fuels or there is the one 8
vendor who can handle up to HEU today or you got a 9
brand new guy coming in that wants to do it all 10 themselves.
11 MR. ALGAMA: Yes, sir.
12 MEMBER PETTI: And that how you would 13 apply these tools would differ for each of those three 14 options, you know, just because of where they are in 15 their licensing basis.
16 MR. ALGAMA: Understood.
17 CHAIR BLEY: Thanks. Anybody else?
18 MS. WEBBER: That's a good comment though, 19 Dave. Thanks for that.
20 MEMBER PETTI: Okay. I mean, one of the 21 things that just it struck me was all of this 22 criticality analysis. Just so you guys are aware, the 23 coaters, where you put the coatings on the particles 24 are critically safe. They're designed to be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 critically safe.
1 So these guys, you know, this is their 2
business, the people who fabricate. They're well 3
aware of all of the rules and incorporating the 4
safety, you know, into the designs of their system.
5 I think it's more difficult when we start 6
talking about the fast reactor fuel, you know, who is 7
going to step forward as an industrial supplier is 8
more difficult. I haven't seen anything, you know, 9
because for years it's just been done, you know, so 10 some, say mom and pop at INL, for the EBI2 core really 11 hasn't been done after that in any large scale.
12 MR. ALGAMA: Yes. I think it's important 13 to understand we're not trying to redo or generate new 14 safety items of interest. We're just trying to find 15 a sufficient number of scenarios that we could test 16 our codes, I think, just so I'm clear. The intention 17 was not to actually do a review. Does that help or?
18 MEMBER PETTI: Yes, I mean, maybe, again, 19 maybe making that clear may be --
20 (Simultaneous speaking.)
21 MEMBER PETTI: -- if it isn't clear enough 22 because that didn't jump out at me, I guess.
23 MR. ALGAMA: Yes, sir. We can make it 24 clear. And doing a full blown review would be a much 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 bigger task that I wasn't anticipating so.
1 MEMBER PETTI: Right.
2 MS. WEBBER: And I think overall, you 3
know, so I reflect on the number of comments related 4
to, you know, scenarios given the breadth of, you 5
know, advance reactor designs. And I think, you know, 6
what common in many of the comments is that we really 7
need to include a set of scenarios, fuel cycle 8
scenarios that will -- I hate to use the word bound, 9
but a set of fuel cycle scenarios that will cover most 10 of what we would anticipate.
11 MR. ALGAMA: Originally, the idea was to 12 do that in the implementation phase. But we can try 13 to hypothesize something up-front but that might 14 change when we start to actually do the work. Is that 15 okay?
16 (Simultaneous speaking.)
17 MR. ALGAMA: I'm sorry. Go ahead.
18 CHAIR BLEY: What I worry about that is if 19 you do it partially now, we've got to make it real 20 clear that it's got to be revisited in substantial 21 detail whether --
22 MR. ALGAMA: Yes, sir.
23 CHAIR BLEY: That's the only answer that 24 I would have.
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66 MR. ALGAMA: We originally thought of 1
giving some more examples of what we would look at.
2 But because of that fear, we decided to keep it just 3
as a plan and then really drill down into it when we 4
implement. But we can try to come up with some 5
compromise approach that makes sense, that provides 6
clarity, if that helps.
7 MS. WEBBER: Well, and I think to -- maybe 8
Dennis, this was your question or maybe it was Joy's 9
question about updating the reports. I mean, this 10 Volume 5 conceivably may be one where given the 11 knowledge that we have today and the uncertainties 12 about where, you know, the fuel cycle technologies are 13 going in the future, especially for the further out, 14 you know, design concepts, this volume may be one that 15 we, you know, note that an update would be necessary 16 potentially.
17 But, you know, I see this document as 18 really providing the strategy. Right now, it contains 19 notionally 10 reports. And, you know, 10 reports and 20 each report represents, you know, a look at that fuel 21 cycle with the identification of gaps and 22 methodologies to close the gaps and, you know, updates 23 to the codes and things like that. But, you know, as 24 we learn more then it may become a set of not only 10 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 but a few others.
1 MR. ALGAMA: Yes, ma'am. It could be 2
bigger or smaller.
3 MS. WEBBER: Right.
4 CHAIR BLEY: That all seems reasonable to 5
me. I WOULD point out to you that although the 6
discussion was about reactors, it applies equally well 7
to fuel cycles.
8 We had a lessons learned letter report 9
recently, a couple other of our letter reports. And 10 in a recent meeting -- actually, I'll go with the OMB, 11 Mr. Fleming, with the group putting together the 12 guidance, where he identified a series of reports in 13 the same vein that lay out approaches to search for 14 initiating events and scenarios for problems.
15 You know, this is people's business where, 16 yes, they're doing it well. But you've got to really 17 do a thorough search to find the things that will 18 surprise or there will be surprises later. So there's 19 some hope for that if you look at those recent 20 references.
21 MR. ALGAMA: Yes, sir. Thank you. You 22 said inside the LMP? I'm sorry.
23 MS. WEBBER: I was going to say, Don, 24 maybe that's something we can talk to Derek and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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68 whoever offline to figure out what those resources are 1
because off the top of my head it doesn't ring a bell.
2 CHAIR BLEY: We can do that. We'll also 3
talk about -- the meeting will be in February. We're 4
on February 20, 21 for Volumes 4 and 5. So we'll have 5
an admin call set up to talk about some of that, and 6
we can give you some of that other information.
7 Anything else from the members? I'm going 8
to go around for public comments and then we'll come 9
back.
10 MEMBER REMPE: Dennis, I guess, again, I 11 would point out that as one searches for initiating 12 events, I think a review of history and root causes 13 for events in the past and what it considers more 14 recent events as well as some of the non-LWR 15 experience in the U.S. where DOE backed the Atomic 16 Energy Commission days where they were the developer 17 as well as the regulator offers some really good 18 lessons in thinking about what needs to be considered 19 here.
20 MR. ALGAMA: Understood.
21 CHAIR BLEY: Can we get the tone line open 22 for comments?
23 MR. DASHIELL: The public bridge line is 24 open for comments.
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69 CHAIR BLEY: Thank you. Is there anyone 1
in the public who would like to make a comment? If 2
so, please state your name and make your comment at 3
this time. Going, going. Okay. We can close the 4
bridge line.
5 Instead of going around to all the 6
members, the intention is to have the meeting in 7
February to write a letter report on Volumes 4 and 5.
8 And I want to divert for just a second back to Kim.
9 Kim, you expressed that you guys didn't have an 10 interest in revisiting the changes to Volumes 1, 2 and 11 3 in the overview report.
12 But I don't know if it fell through the 13 cracks, or crack, because of COVID or if there's other 14 reasons, but we have never received any real response 15 letter on our letter on Volumes 1, 2 and 3. So given 16 that we hadn't --
17 MS. WEBBER: Actually, I have that. I 18 think I have that because I think we crafted it. But 19 I think we can try to dredge that up.
20 CHAIR BLEY: That might take care of any 21 revisiting them in February. So if you can find that 22 and get it in the system, we'll talk about that, too, 23 when we put them up. I'd like to revisit those 24 because so far we don't have anything from you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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70 officially.
1 MS. WEBBER: Okay. Yes. I'll see if I 2
can resurrect that. But I think I recall, you know, 3
there was a specific ticket with a response.
4 CHAIR BLEY: And it never made it up on 5
the NRC website either, it's normally there.
6 MS. WEBBER: Okay.
7 CHAIR BLEY: So the intention is to write 8
a letter on Volumes 4 and 5 and maybe it's something 9
about dealing with our previous recommendations from 10 November of last year.
11 Are there any members of the subcommittee 12 at this time who would like to comment specifically?
13 Instead of going all around the room, I'll just ask 14 you to come forward. Mike Corradini, anything from 15 you as our consultant?
16 MR. WIDMAYER: Hey, Dennis, this is Derek.
17 Mike's currently out of the meeting.
18 CHAIR BLEY: Oh, okay. He said he might 19 not be here. I saw him so I screwed up one. Okay.
20 So without any further comments, we'll look forward to 21 getting together in February to talk about Volumes 4 22 and 5. We'll have that offline meeting with Kim and 23 maybe some others before then. So at this time, we 24 are adjourned.
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71 MR. ALGAMA: Yes, sir. Thank you very 1
much.
2 MS. WEBBER: Yes. Hey, Dennis, is there 3
a date for that fall committee meeting?
4 CHAIR BLEY: Oh, geez, Derek? Yes, it's 5
in February.
6 MR. WIDMAYER: Yes. We have dates but we 7
haven't done an agenda or anything yet but.
8 CHAIR BLEY: We don't have it pinned down.
9 It will be the 4th or the 5th.
10 MR. WIDMAYER: Yes.
11 MS. WEBBER: Oh, okay. That's good enough 12 for now.
13 MR. WIDMAYER: Yes.
14 MS. WEBBER: Okay. Right. Well, I do 15 appreciate you all taking the time and putting some 16 really good thoughts together about how to improve not 17 only the strategy but the quality of the report. And 18 I just really appreciate your time. I know you're 19 busy, and there's a lot going on. So thank you very 20 much.
21 MR. LEE: This is Richard Lee. I want to 22 make a comment.
23 MS. WEBBER: Okay.
24 CHAIR BLEY: Okay. I guess we can reopen 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 and take your comment.
1 MR. LEE: In response to Dennis, I mean, 2
Dave Petti about the fast reactor fuel fabrication, 3
our staff can reach out to the French and the Japanese 4
to learn what they have done with respect to the fast 5
reactor stuff so.
6 MEMBER PETTI: Yes. But, Richard, that's 7
oxide fuel. And the U.S. is the only ones who make 8
the metal fuel.
9 MR. LEE: Yes, but the thing is that you 10 are worried about mostly, like, the enrichment aspect 11 of it. So there may be some applicability from those.
12 MEMBER PETTI: That's true, yes.
13 MR. LEE: Yes.
14 MEMBER KIRCHNER: Yes, that part might be.
15 But as Dave points out -- this is Walt Kirchner. Yes, 16 their experience is mainly oxide. We had at that TF 17 oxide fuel. But the concepts that we see coming seem 18 to be leaning towards using the metallic fuel, which 19 is the argon INL EBR-II experience.
20 MR. LEE: Let us remember if I'm going to 21 validate the neutronics aspect of it, I can use a lot 22 of different forms in terms of criticality so. The 23 physics is still there with fast spectrum behavior for 24 the uranium aspect of it.
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73 MR. MOORE: Chairman Bley, this is Scott 1
Moore. Can I be recognized?
2 CHAIR BLEY: Yes, you may, Scott. Go 3
ahead.
4 MR.
MOORE:
To follow-up on the 5
conversation, the full committee meeting in February 6
is on February 4 and 5. And as Derek mentioned, it 7
does not yet have an agenda.
8 The second thing is just to note that 9
Steve Schultz is also in the meeting or at least the 10 list of attendees is showing Steve, our consultant on.
11 CHAIR BLEY: Thank you very much.
12 MS. CUBBAGE: Dr. Bley, this is Amy 13 Cubbage. May I be recognized?
14 CHAIR BLEY: Who is this?
15 MS. CUBBAGE: Amy Cubbage.
16 CHAIR BLEY: Yes, Amy.
17 MS. CUBBAGE: Yes, I just wanted to note 18 that the staff contracted with the national labs to 19 look at the safety and hazards associated with fuel 20 fabrication in the reports available on the NRC 21
- website, including specifically a
metal fuel 22 fabrication safety hazards report.
23 CHAIR BLEY: Thank you. And that's 24 publicly available now?
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74 MS. CUBBAGE: Yes, it is. I can provide 1
the link to Derek.
2 CHAIR BLEY: Thank you. That will be 3
helpful. Well, we sort of reopened the meeting. I 4
think I heard Joy.
5 MS. WEBBER: No, it was Kim. Amy, can you 6
copy me on that, too?
7 MS. CUBBAGE: Absolutely.
8 MS. WEBBER: Thank you.
9 CHAIR BLEY: Anybody else? We're 10 finishing way early. I already thought we were 11 adjourned once, but I'll give you another minute here.
12 Okay. If nothing more, we will adjourn at 13 this time for real. And we'll see you again in 14 February. Thanks to all.
15 MS. WEBBER: Thank you all. Happy 16 Holidays.
17 CHAIR BLEY: Happy holidays. Bye-bye.
18 MR. ALGAMA: Thank you. Goodbye.
19 (Whereupon, the above-entitled matter went 20 off the record at 10:59 a.m.)
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Implementation Action Plan (IAP)
Strategy 2 - Volume 5 Code Application Plans for Advanced Reactor Nuclear Fuel Cycles December 1, 2020 Kimberly A. Webber, Ph.D.
Division of Systems Analysis Office of Nuclear Regulatory Research
Agenda
- Staff Introduction
- IAP Strategy 2 Overview
- ACRS Strategy 2 Meeting Schedule
- Non-LWR Fuel Cycle Analysis Plan (Vol. 5)
- Overview of Existing Fuel Cycle and Analysis
- Advanced Reactor Fuel Cycle and Analysis
- Leveraged Programs
- Concluding Remarks 2
- Improve mission value while enabling safe operations
- Deliver cost savings
- Develop regulatory tools
- Leverage collaborations
- Build staff expertise NRCs Be Ready Attitude 3
BlueCRAB
NRCs Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Knowledge, Skills, and Capacity Strategy 2 Analytical Tools Strategy 3 Flexible Review Process Strategy 4 Industry Codes and Standards Strategy 5 Technology Inclusive Issues Strategy 6 Communication ML17165A069 4
Introduction ML20030A174 Volume 1 ML20030A176 Volume 3 ML20030A178 Volume 2 ML20030A177 These Volumes outline the specific analytical tools to enable independent analysis of non-LWRs, gaps in code capabilities and data, V&V needs and code development tasks.
Strategy 2: Computer Code Readiness Code Development Plans Volume 4 ML20028F255 Volume 5 ML20308A744 5
NRCs Integrated Action Plan (IAP) Status 6
Overview of Volume 5 Assessment and use of existing NRC computational tools for accident analysis (Volume 3) and consequences (Volumes 3/4)
Incremental development approach based on existing LWR fuel cycle as reference Staff experience with anticipated non-LWR fuel cycle and use of computer codes Development of non-LWR fuel cycle reports and publicly available input decks Volume 5 ML20308A744 7
NRC non-Light Water Reactor Vision and Strategy, Volume 5: Radionuclide Characterization, Criticality, Shielding, and Transport in the Nuclear Fuel Cycle Presented by Don Algama (RES) and Drew Barto (NMSS)
United States Nuclear Regulatory Commission Office of Nuclear Regulatory Research (RES)
Nuclear Materials Safety and Safeguards (NMSS) 1
Acknowledgements
- Dr. David Luxat (Sandia) and Dr. William Wieselquist (ORNL) were instrumental in the plan development.
2
IAP Strategy 2 Volumes to Date 3
ML20030A177 ML20030A174 ML20308A744 ML20030A176 ML20030A178 ML20028F255 Introduction Volume 1 Volume 2 Volume 3 Volume 4 Volume 5
Objectives
- Elements of the fuel cycle plan
- Demonstrate computer code readiness
- Assessment and use of existing NRC computational tools for accident analysis (Volume 3) and consequences (Volumes 3/4)
- Incremental development approach based on existing LWR fuel cycle as reference
- Staff experience with anticipated non-LWR fuel cycle and use of computer codes
- Development of non-LWR fuel cycle reports and publicly available input decks 4
Regulatory Application of Codes 5
Reactor Aux. Systems; Transport; Storage Engineered Safety Features Fuel Mechanical Design Nuclear Design Thermal-Fluid Design GDC 2 GDC 10 GDC 10, 11, 12, 26, 27, 28 GDC 10 GDC 12 Dynamic Stress Analysis Thermo-Mechanical Performance Reactor Physics Analysis Systems Analysis/Thermal Margin Stability Analysis FAST SCALE/PARCS/TRACE Criticality/Shielding 10 CFR 50.68 (GCD 62); 10 CFR 70, 71 and 72 Materials Accounting Criticality and Shielding Analyses Inventory SCALE 10 CFR 74 Containment Core Cooling 10 CFR 50.44 GDC 16, 38, 50 10 CFR 50.46 GDC 10, 34, 35 Safety Review 10 CFR 50.34; 10 CFR 50.67 10 CFR 52.47; 10 CFR 100 GDC 19 10 CFR 50.160 (EP) 10 CFR 52.17 (ESP) 10 CFR 52.79 (COL)
Protection Against Radiological Release Environmental Review 10 CFR 51.30 10 CFR 51.50 10 CFR 51.70 10 CFR 51.75 Containment Analysis Systems Analysis (inputs from fuel thermo-mechanical and reactor physics analyses)
Source Term/Dose Consequence Analysis MELCOR
- FAST, SCALE/TRACE/PARCS MELCOR/
MACCS RADTRAD/
RASCAL SCALE
Transportation and Storage Licensing (LWR)
ORIGAMI Reactor-specific radioactive isotopics/source term characterization AMPX Validated cross section libraries in multigroup (O(100g)) or continuous-energy (O(100,000g);
depletion and decay data ENDF/B Physics data Thermal scattering law, resonance data, energy distributions, fission yields, decay constants, etc.
CSAS 3D criticality safety analysis SHIFT/MAVRIC 3D shielding and dose rate analysis JEFF Activation Isomeric cross sections, activation reactions Sources4C neutron emission data (alpha,n)
TRITON/SHIFT General reactor fuel neutron transport +
depletion ICRP dose conversion factors, radiotoxicity NIST natural abundance, atomic mass ORIGEN General depletion, decay, source term analysis end-points 6
Thermal Analysis Structural and containment (cask) analyses
Severe Accident & Consequence Analysis (LWR/non-LWR example)
ORIGAMI Reactor-specific radioactive isotopics/source term characterization AMPX Validated cross section libraries in multigroup (O(100g)) or continuous-energy (O(100,000g); depletion and decay data ENDF/B Physics data Thermal scattering law, resonance data, energy distributions, fission yields, decay constants, etc.
JEFF Activation Isomeric cross sections, activation reactions Sources4C neutron emission data (alpha,n)
TRITON/SHIFT General reactor fuel neutron transport + depletion NIST natural abundance, atomic mass ORIGEN General depletion, decay, source term analysis end-points Kinetics Data nuclide-specific beta-effective, precursor data MACCS Offsite consequence analysis MELCOR Severe accident progression and mechanistic source terms NRC Non-Light Water Reactor Vision and Strategy, Volume 3 - Computer Code Development Plans for Severe Accident Progression, Source Term, and Consequence Analysis, Revision 1, January 2020, ML20030A178 7
Examples of Existing Fuel Cycle Analysis
- Level 3 PRA Project
- SCALE/MELCOR are used to support PRA development of accident sequences and source terms including non-reactor scenarios for the spent fuel pool
- SCALE/MELCOR was used to study the performance of a SFP under severe accident conditions
- NUREG/CR-7108/7109
- Here SCALE was used to estimate isotopic depletion and criticality code, and cross section data bias related to burnup credit in spent fuel storage and transportation systems 8
Examples of Existing Fuel Cycle Analysis Barnwell - Non-Reactor Safety Assessment SCALE/MELCOR utilized as part of best-estimate analysis methodology in NUREG/CR-7266 Spent fuel inventories developed in SCALE package Aerosol transport modeling Integral analyses estimate radiological transport and release Aerosol modeling enables estimation of transport of hazardous material within facility and to environment Accident scenarios considered relevant to broad range of facility accidents Explosion scenario Fire scenario Combined explosion and fire scenario 9
non-LWR Characteristics 10 Table 1-1. Comparison Between LWR and Non-LWR Reactor Type Enrichment (wt.%)
Fuel Form Typical Discharge Burnup Fuel Residence Time On-Site Fuel Processing Fuel Storage /
Transport LWR (Ref.)
<5 U Oxide Peak Rod Average:
<62 GWd/MTU Max Assembly Average:
<55 GWd/MTU Assemblies burned for approximately 3 to 4 cycles No Storage:
Fresh and spent fuel storage on-site or off-site Transport:
FE: UF6 solid transport in 30B cylinders, fresh fuel assembly and fuel component (UO2 powder/pellet) transportation packages BE: Used fuel transport and dry storage containers LWR: HALEU
/HBU (Ref.)
5 - 10 U Oxide Peak Rod Average:
~75 Wd/MTU Max Assembly Average:
~60-70 GWd/MTU Assemblies burned for approximately 3 to 4 cycles No HPR 5 - 20 U Oxide U Metal 2-10 GWd/MTU Up to 7yrs No To be evaluated*
SFR 5 - 20 U Metal Up to 300 GWd/MTU To be evaluated*
No To be evaluated*
HTGR 5 - 20 TRISO (UCO or UO2) in pebble bed or prismatic array 100-200 GWd/MTU To be evaluated*
No To be evaluated*
FHR 5 - 20 TRISO (UCO or UO2) in pebble bed 100-200 GWd/MTU To be evaluated*
No To be evaluated*
MSR 5 - 20 235U dissolved in molten salt To be evaluated 2-3yrs Yes To be evaluated*
- 1 atom-% burnup is approximately 9.4 GWd/MTU.
- Will be evaluated based on information available at the time work is undertaken, e.g. based on current DOE and industry input.
Analysis Approach Develop accident scenarios by reviewing available information including documents such as:
- NUREG/CR-6410 Nuclear Fuel Cycle Facility Accident Accident Analysis Handbook
- NUREG-1520 Standard Review Plan for Fuel Cycle Facilities License Applications
- NUREG-2215 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities - Final Report
- NUREG-2216, Standard Review Plan for Spent Fuel Transportation
- DOE-HDBK-1224-2018: DOE Accident Analysis Handbook Hazard and Accident Analysis Handbook 11
Scope of Analysis Assess existing codes to cover neutronics and radionuclide and non-radionuclide hazards throughout non-LWR fuel cycles Consequence and radiation protection methods are covered under Volume 3/4 Mining, milling, long term storage and disposal are not considered in this activity Leverage volume 3 non-LWR designs
- Fluoride-Salt-Cooled (Solid-Fuel) High Temperature Reactor (FHR)
- Heat Pipe Reactors (HPR)
- Sodium Fast Reactor (SFR)
- High Temperature Gas Reactor (HTGR)
- Molten Salt Reactor (MSR) 12 Follow these analysis steps used in Volume 3 and previous fuel cycle work for LWRs 1.
Define scenario 2.
Identify safety related item(s) of interest 3.
Ask the right safety questions /
Phenomena of interest / Understand the dominant features 4.
Survey experiments available that provide fundamental information 5.
Develop physics models to capture dominant feature and allow prediction 6.
Translate physics models into computer code 7.
Perform verification testing (unit testing; and integrated testing as code complexity increases) 8.
Perform validation with experiments.
Capture the integrated codes performance (with uncertainty analysis) 9.
Document findings
Deliverables 10 reports are defined as a result of this plan Each report defines a set of accident scenarios during a portion of the fuel cycle Perform assessment, analysis, and generate demonstration input files 5 non-LWRs currently considered and openly available reference designs defined in volume 3:
1.
FHR Fuel Cycle Analysis (Berkeley Mk. 1) 2.
HPR Fuel Cycle Analysis (INL Design A-MET) 3.
SFR Fuel Cycle Analysis (MET-1000/VTR) 4.
HTGR Fuel Cycle Analysis (PBMR-400) 5.
MSR Fuel Cycle Analysis (MSRE) 5 front end (FE) reports centralize FE analysis among these non-LWRs 6.
Enrichment and UF6 Handling up to 20 wt.%
7.
TRISO Fuel Kernel Fabrication 8.
Uranium Metallic Fuel Fabrication 9.
Fast Reactor Fuel Assembly Fabrication
- 10. Pebble TRISO Fuel Fabrication 13 This organization of deliverables allows prioritizing specific designs and reducing overlap. For example:
HTGR analysis requires the following reports 67104.
For FHR, it would require 67101. 6,7, and 10 are already available!
Reference - LWR Cycle 14 Each analysis report tackles one or more of the equivalent fuel cycle stages for each non-LWR.
NOTE: Transportation off-site and off-site storage (T3 and S1) are currently not considered in this fuel cycle assessment plan due to uncertainty with this part of the back end.
FHR Fuel Cycle Report 15 The FHR fuel cycle report develops and analyzes new accident scenarios related to stages U1 and U4 and links them to earlier front-end stages (E1, T1, F1, F2, T2) analyzed in this project and in-reactor scenarios U2 from volume 3.
HPR Fuel Cycle Report 16 The HPR fuel cycle report develops and analyzes new accident scenarios related to stages F2, T2, U1 and U4 but also requires re-analysis of U2 for a metallic fuel system (current source term demo calcs using oxidic fuel). NOTE: The F2 and T2 front end stages are included in this report because fabrication and transportation of an HPR core will be specific to that design and thus nothing is gained from putting those stages in their own analysis reports.
SFR Fuel Cycle Report 17 The SFR fuel cycle report develops and analyzes new accident scenarios related to stages U1, U3, and U4 and links them to previously studied E1, T1, F1, F2, and T2. NOTE: The F2 and T2 front end stages are their own report not because of overlap included in this report because fabrication and transportation of an HPR core will be specific to that design and thus nothing is gained from putting those stages in their own analysis reports.
HTGR Fuel Cycle Report 18 The HTGR fuel cycle report develops and analyzes new accident scenarios related to stages U1 and U4 and links them to front-end stages (E1, T1, F1, F2, T2) analyzed in this project and in-reactor accident scenarios U2 from volume 3.
Front end analysis is basically the same as for FHR.
MSR Fuel Cycle Report 19 The MSR fuel cycle report has the least overlap with any other design and develops and analyzes new accident scenarios for F1, T2, U1, and U4 in the main MSR analysis and links them only to front end E1 and T1 for UF6 enrichment and transportation.
Leveraged Programs
- UF6 transport packages
- Fresh fuel transport packages
- Volume 3 (codes and plant models)
- Capabilities to characterize utilization stage
- Hazardous material transport for non-water systems
- DOE Programs
- DOE-NE spent fuel and waste science and technology program
- Support hazard identification and characterization 20
Concluding Remarks
- Relying on a reasonable and flexible approach
- Sufficient capabilities to support non-LWR fuel cycle analyses
- Decades of model development and validation can be applied to non-LWR analyses as in Volume 3 and other programs
- Plan will be updated as more experience is gained and as new information becomes available 21