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