ML22060A171
| ML22060A171 | |
| Person / Time | |
|---|---|
| Issue date: | 02/17/2022 |
| From: | Michael Snodderly Advisory Committee on Reactor Safeguards |
| To: | |
| Snodderly, M, ACRS | |
| References | |
| NRC-1859 | |
| Download: ML22060A171 (395) | |
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: Thursday, February 17, 2022 Work Order No.: NRC-1859 Pages 1-251 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (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.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
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 THURSDAY 10 FEBRUARY 17, 2022 11 + + + + +
12 The Subcommittee met via Teleconference, 13 at 9:30 a.m. EST, David A. Petti, Chairman, presiding.
14 15 COMMITTEE MEMBERS:
16 DAVID A. PETTI, Chairman 17 RONALD G. BALLINGER, Member 18 VICKI M. BIER, Member 19 CHARLES H. BROWN, JR. Member 20 VESNA B. DIMITRIJEVIC, Member 21 GREGORY H. HALNON, Member 22 WALTER L. KIRCHNER, Member 23 JOSE MARCH-LEUBA, Chairman 24 JOY L. REMPE, Member 25 MATTHEW W. SUNSERI, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 1 ACRS CONSULTANTS:
2 STEPHEN SCHULTZ 3 DENNIS BLEY 4
5 DESIGNATED FEDERAL OFFICIAL:
6 MIKE SNODDERLY 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 1 C-O-N-T-E-N-T-S 2 PAGE 3 Opening Remarks . . . . . . . . . . . . . . . . . 9 4 Staff Introduction . . . . . . . . . . . . . . . 14 5 History and Evolution of LWR Source Term . . . . 14 6 NRC analytical tools and past studies . . . . . . 40 7 Light Water SMR Design Certification 8 Source Term Approach and Source Term 9 Approach for Early Non-LWR Movers . . . . . . . . 70 10 Michelle Hart, NRR . . . . . . . . . . . . 70 11 Accident Consequence-Related Regulation 12 Activities . . . . . . . . . . . . . . . . . . 116 13 Guidance and Information for Developing 14 Advanced Reactor Source Terms . . . . . . . . . 123 15 Guidance for Developing Advanced Reactor 16 Source Term (Long Term) . . . . . . . . . . . . 181 17 Questions and Comments . . . . . . . . . . . . 190 18 Opportunity for Public Comment . . . . . . . . 202 19 Member Discussion . . . . . . . . . . . . . . . 211 20 Adjourn . . . . . . . . . . . . . . . . . . . . 251 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 1 P R O C E E D I N G S 2 9:30 a.m.
3 CHAIRMAN PETTI: Okay, committee members, 4 interested staff, and stakeholders. The meeting will 5 come to order.
6 This is a meeting of the Advisory 7 Committee, on the Active Safeguards Future Plant 8 Designs Subcommittee.
9 I'm Dave Petti, lead member for the 10 meeting.
11 Members in attendance today are Ron 12 Ballinger, Vicki Bier, Greg Halnon, Jose March-Leuba, 13 Matt Sunseri, Walt Kirchner, Charlie Brown, consultant 14 Dennis Bley is on with us as well.
15 I do not see Steve Schultz yet, but I do 16 expect him. And, I do not see Member Dimitrijevic.
17 MEMBER DIMITRIJEVIC: I'm here.
18 CHAIRMAN PETTI: Mike Snodderly is the 19 designated federal --
20 (Simultaneous speaking.)
21 MEMBER DIMITRIJEVIC: I'm attending.
22 CHAIRMAN PETTI: Oh, thank you, Vesna.
23 MEMBER DIMITRIJEVIC: I don't know why you 24 don't see me. They didn't see me yesterday, but I'm 25 here.
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5 1 CHAIRMAN PETTI: Ahh, great.
2 DR. SCHULTZ: This is Steve, Dave. I'm 3 here.
4 CHAIRMAN PETTI: Okay, good. And, 5 consultant Steve Schultz is here. Great.
6 Mike Snodderly is the designated federal 7 official for this meeting.
8 The subcommittee will discuss with the 9 staff the Integration of their Source Term Activities 10 in Support of Advanced Reactor Initiatives.
11 This is a topic that is of interest both 12 to the committee, and the staff. It's nice to see 13 minds coming together and thinking that this was an 14 important topic for us to talk, about in the context 15 of Part 53.
16 The ACRS was established by statute, and 17 is governed by the Federal Advisory Committee Act, 18 FACA.
19 The NRC implements FACA in accordance with 20 its regulations found in Title 10 of the Code of 21 Federal Regulations, Part 7.
22 The committee can only speak through its 23 published letter reports. We hold meetings to gather 24 information, and perform preparatory work that will 25 support our deliberations at a full committee meeting.
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6 1 The rules for participation in all ACRS 2 meetings were announced in the Federal Register, on 3 June 13, 2019.
4 The ACRS section of the U.S. NRC public 5 website provides our chat or bylaws agendas, letter 6 reports, and full transcripts, of all full and 7 subcommittee meetings, including slides presented 8 there.
9 The agenda for this meeting was posted 10 there as well.
11 As stated in the Federal Register notice, 12 members of the public who desire to provide written or 13 oral input to the subcommittee may do so, and should 14 contact the designated federal official five days 15 prior to the meeting, as practicable.
16 This is an MS Teams virtual meeting, and 17 members of the public may listen in on the 18 presentations, and committee discussion using the call 19 in number and the conference ID number included on the 20 agenda.
21 We have received no written (Audio 22 interference) to make oral statements from members of 23 the public regarding today's meeting.
24 There will be an opportunity for public 25 comment, and we set aside 10 minutes in the agenda for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 1 comments from members of the public attending, or 2 listening to our meetings.
3 Written comments may be forwarded to Mike 4 Snodderly, the designated federal official.
5 A transcript of the open portions of the 6 meeting is being kept, and it is requested that the 7 speakers identify themselves, and speak with 8 sufficient (Audio interference) so that they can be 9 readily heard.
10 Additionally, participants should mute 11 themselves when not speaking.
12 I just want to note that the staff has put 13 together a tremendous amount of technical information 14 to convey to the committee today, in, you know, 15 basically our one day meeting.
16 I'm going to be extra mindful of the 17 schedule, to make sure we get through all the 18 information before the end of the day.
19 Now, I'd like to have Larry Burkhart 20 provide a comment about MS Teams participation, 21 followed by Arlon, who will provide us with an 22 overview of the agenda.
23 And, then John Segala will make an opening 24 statement (Audio interference) Larry.
25 MR. BURKHART: Okay, thanks. Just for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 1 everyone's notice, the ACRS is now providing the MS 2 Teams link for all open meetings, to all members of 3 the public.
4 I do notice that we have a couple folks 5 that are tying in via phone. So, if there are any of 6 the public attendees who need the MS Teams link, 7 please email me at the following address: L-A-W-R-E-N-8 C-E, Lawrence, dot Burkhart, B-U-R-K-H-A-R-T, 9 @NRC.GOV.
10 So again one more time, if any members of 11 the public who don't have the MS Teams link would like 12 it, please email me at lawrence.burkhart@NRC.gov.
13 Thanks.
14 MR. ACOSTA: Okay, this is Arlon Acosta.
15 I just would like to go over quickly, over the agenda.
16 We're going to as Dave already said, going 17 to have the opening remarks from John, and staff 18 introduction.
19 Following him, we will continue with the 20 presentations as follows: the History and Evolution of 21 LWRS Source Terms; in the NRC analytical tools and 22 past studies; SCALE and MELCOR in non-light water 23 reactors reference analysis.
24 And, then somewhere in between that 25 presentation, we will have a break that Dave Petti NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 1 will tell us.
2 And, we may need to continue with that 3 item, which is item 4 is SCALE and MELCOR non-light 4 water reactor reference plant analysis.
5 And, then subsequently, we will continue 6 with the NuScale EPZ Sizing Methodology Topical Report 7 Revision 2 presentation; light water reactor SMR 8 design certification source term approach; source term 9 approach for the early non-light water movers.
10 And then sometime as designated by Dave, 11 we'll have a lunch break, and Accident and followed by 12 the other presentation on Accident-Consequence-related 13 regulatory, regulation activities.
14 And, again, place to be determined for a 15 break, and four more, two more presentations on 16 guidance and information for developing advance 17 reactor source term.
18 And, guidance for developing advanced 19 reactor source term as far as long-term is concerned.
20 And, as Dave mentioned, an opportunity for 21 public comment and member discussion, and subsequently 22 to adjourn.
23 John Segala?
24 MR. SEGALA: Thank you, Arlon. Hopefully 25 you can hear me.
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10 1 I'm John Segala, Acting Deputy Director of 2 the Division of Advanced Reactors and Non-Power 3 Production Utilization Facilities in the Office of 4 Nuclear Reactor Regulation.
5 We're pleased to be here today for this 6 important topic. Determining source term is a 7 critical component in NRC's licensing of advanced 8 reactor designs.
9 Over the past few years, we have discussed 10 source term with the ACRS in a variety of contexts, 11 including emergency preparedness rulemaking, licensing 12 modernization project, Part 53 analytical tool 13 development, the NuScale review, et cetera.
14 The ACRS has written several letters and 15 raised a number of issues in this area, that 16 development of a design specific source term will 17 require substantial work.
18 The new reactor designers may need 19 expanded guidance in this area, and this could help 20 make the reviews more efficient. And, staff efforts 21 need to be coordinated across the various source term 22 related activities.
23 So, we proposed to have this separate ACRS 24 meeting today, to focus on source terms to address 25 these comments in a more holistic and coordinated way.
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11 1 We understand that ACRS may want to write 2 a letter on this topic, and have tentatively scheduled 3 a discussion on this topic during the March 2nd ACRS 4 full committee meeting.
5 Source term has evolved over the past 60 6 or so years. Historically for large light water 7 reactors that are operating today, the NRC developed 8 a source term for calculating offsite doses for 9 citing, and provided it to the reactor designers.
10 For example, the TID-14844 are source 11 term.
12 To facilitate the use of a risk informed 13 performance based licensing approach such as the 14 licensing modernization project, design and/or 15 scenario specific source terms will need to be 16 developed by the applicants.
17 With respect to guidance for source term 18 development, the NRC staff has completed a number of 19 source term related projects, and more work in 20 ongoing. All of which we will be presenting to the 21 ACRS today.
22 Although we acknowledge that developing 23 design and scenario specific source term will require 24 substantial work, the staff believes that there is 25 sufficient guidance available for applicants to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
12 1 develop design and scenario specific source terms, and 2 for the NRC staff to complete its technical review.
3 We will consider the need for additional 4 source term guidance in the future, based on our 5 experience with the ongoing pre-application and 6 licensing interactions that we have, that we are 7 having with applicants over the next couple years.
8 Such as Kairos, X-energy, TerraPower.
9 We have recently developed a new public 10 webpage focused on source term for nuclear power 11 reactors, which contains information on source term, 12 and provides links to a number of source term guidance 13 documents, videos, presentations, and other associated 14 references such as SECY papers, staff requirements, 15 memorandums, NUREGs, contractor reports.
16 We plan to showcase the webpage for you 17 today.
18 In addition to this available guidance and 19 information on source term, the staff encourages 20 applicants to engage in pre-application activities, to 21 seek early NRC feedback on their source term 22 methodology.
23 And, we have been having such engagements 24 with several pre-applicants, which we will discuss 25 today.
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13 1 Pre-application engagement on topics such 2 as source term, will support a more efficient 3 application review.
4 If you could go to the next slide?
5 Thanks.
6 At this point, I'd like to introduce the 7 NRC team that will be making the presentations today.
8 We have four NRC staff from the Division of Advanced 9 Reactors and Non-Power Production and Utilization 10 Facilities, in the Office of Nuclear Reactor 11 Regulation.
12 That includes Michelle Hart, Jason 13 Schaperow, Bill Reckley, Tim Drzewiecki. We also have 14 Mark Blumberg, from the Division of Risk Assessment in 15 the Office of Nuclear Reactor Regulation, and we have 16 Hossein Esmaili, in the Division of Systems Analysis, 17 in the Office of Nuclear Regulatory Research.
18 We are looking forward to hearing from the 19 ACRS today on source term, and any insights and 20 feedback you may all have.
21 That completes my opening remarks.
22 MR. BLUMBERG: Thank you, John. My name is 23 Mark Blumberg, and I'll be giving you a presentation, 24 which is a brief overview of the historical use, and 25 the evolution of light water reactor source terms in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 1 the NRC regulatory process.
2 The NRC exists to protect the health and 3 safety of the public from the accidental release of 4 fission products.
5 One of the ways the NRC staff and 6 licensees determine what measures and barriers are 7 needed, is to perform dose analyses.
8 Specifically, these conservative analyses 9 address the situation where we were wrong about the 10 success of a facility's response to events or 11 accidents.
12 Dose analyses provide an effective way to 13 account for the uncertainties in equipment, and human 14 performance.
15 In particular, these analyses account for 16 the unlikely events that involve unknown, or 17 unforeseen failure mechanisms, or phenomena, which 18 because they are unknown or unforeseen, are not 19 reflected in the PRA or traditional engineering 20 analyses.
21 A critical component of the dose analysis 22 is the source term. In the Regulations for Part 50, 23 the NRC defines the source term as the magnitude and 24 mix of the radionuclides released from the fuel, 25 expressed as fractions of fission product inventory in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 1 the fuel, as well as the physical and chemical form 2 and timing of their release.
3 Please go to slide 6 now.
4 Since we're talking about history in this 5 presentation, rather than providing an outline, a 6 timeline is presented in this slide instead.
7 This slide presents a timeline of the 8 documents and events, surrounding the development of 9 the light water reactor source terms.
10 Since we are providing an overview today, 11 this timeline highlights a few of the more important 12 documents and events pertinent to our regulatory 13 approach.
14 The text in the boxes are color coded.
15 Boxes with text written in black describe important 16 documents such as the NUREG-1465 document, shown here 17 on the timeline in 1995.
18 NUREG-1465 supports our most recent source 19 term and regulatory guidance.
20 Boxes with the text written in red 21 describe the important events. The events are located 22 below the timeline, and provide the timeframe for the 23 first critical pile, the 1954 Atomic Energy Act that 24 allowed civilian nuclear power development, and four 25 accidents that influenced the development of our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 1 regulatory source terms.
2 I'll discuss many of these documents and 3 events in this presentation, and this timeline can be 4 used to provide a perspective of the timing of these 5 events and documents.
6 Go to slide 7, please.
7 In the 1940s and the 1950s, the first 8 government reactors were placed far from population 9 centers, and therefore, siting was not a critical 10 issue.
11 Over time when the utilities wanted to use 12 reactors for power, transmission costs were a 13 consideration. Therefore, utilities sought to put 14 reactors closer to population centers that they 15 served.
16 Because of their potential hazard that 17 reactors posed, reactors with containments were 18 proposed to provide defense and depth against 19 accidents.
20 These reactor containments provide an 21 extra barrier to prevent radioactivity from reaching 22 the environment, in the unlikely scenario that the 23 reactor coolant system failed, and the fuel melted.
24 To determine the acceptability of these 25 designs, at the time the Atomic Energy Commission NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 1 proposed using population densities to site reactors.
2 But public opinion judged this method was 3 too rigid. The AEC, the Atomic Energy Commission, and 4 the industry, ultimately decided that siting would be 5 performed on a case-by-case basis, using dose 6 calculations.
7 Please go to slide 8.
8 In 1962, the NRC issued 10 CFR Part 100.
9 This performance based rule used dose calculations to 10 evaluate the defense and depth provided by 11 containment.
12 Nearly all the current operating reactors 13 were licensed originally to 10 CFR Part 100, and TID-14 14844, which, as John said, provide guidance on a 15 containment source term to be used for LOCAs involving 16 significant fuel melt.
17 This source term was based upon heating 18 fuel chips in a furnace and seeing what was released.
19 The source term involves the release of 20 100 percent of the Noble gasses, 50 percent of the 21 iodine, and 1 percent of the other radionuclides as 22 particles.
23 The iodine released is mostly elemental 24 iodine. The remaining iodine is assumed to be 25 particulate, and organic.
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18 1 This deterministic source term is assumed 2 to be available instantaneously, at the start of the 3 accident.
4 Accident models credit mitigation of this 5 source term by safety related systems, structures, and 6 components. Non-LOCA source terms are also provided 7 in Reg Guide 1.195.
8 You can go to slide 9 now, please.
9 Source term estimates under core melt 10 conditions became of great interest shortly after the 11 Three Mile Island, or TMI accident, when it was 12 observed that only a relatively small amount of iodine 13 was released to the environment.
14 Although the release of iodine in the 15 containment may have been close to the TID source 16 term, the releases to the environment were much 17 smaller than suggested in regulatory models.
18 In 1981, the NRC began a major research 19 effort to obtain a better understanding of the fission 20 product transport and release mechanisms in LWRs, 21 under these severe accidents.
22 This effort involved several national 23 laboratories, extensive NRC staff in the NRC, and 24 nuclear industry groups.
25 The cooperative research resulted in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 1 development of the source term copackage to examine 2 core melt progression, and fission product release and 3 transport in LWRs.
4 In order to determine the accident source 5 terms for regulatory purposes, a range of severe 6 accidents were examined.
7 This work was based upon the work done in 8 NUREG-1150. It provided an assessment of severe 9 accident risks in five LWRs.
10 In addition, some source term code packets 11 calculations and insights from the MELPOR code, were 12 used.
13 These efforts provided research that 14 confirmed that the source term release is highly 15 dependent upon the nature of the accident, which 16 included the accident pressures, temperatures, and 17 release pathways.
18 If you want to go to slide 10 now, please.
19 In December of 1999, the NRC issued the 20 final rule known as 10 CFR 5067. This rule is also 21 known as the alternative source term, or AST rule.
22 It allowed plants that were licensed under 23 10 CFR Part 100, to convert to the AST, 10 AST. An 24 acceptable AST based upon NUREG-1465 was provided in 25 a document known as Reg Guide 1.183.
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20 1 Based upon Commission direction, this AST 2 used only the releases specified in NUREG-1465 up to 3 the early end vessel release phase.
4 Commensurate with fuel utilization at the 5 time, the NUREG-1465 source term was limited to core 6 average burn ups of 40-gigawatt days per metric ton 7 uranium, and five weight percent enriched uranium.
8 The AST rule was written to be flexible 9 though so it did not specify in AST, a specific AST.
10 Alternatives could be proposed and used, as technology 11 and fuel utilization changed over time.
12 To facilitate this change, Reg Guide 1.183 13 included a list of significant attributes for future 14 ASTs.
15 And, go to slide number 11 now, please.
16 The LOCA source term in Reg Guide 1.183, 17 is specific to whether the reactor is a BWR or a PWR.
18 This table provides a comparison of the BWR source 19 term to the TID-14844 source term.
20 Significant differences in the release 21 timing, release fractions, and chemical form of the 22 source term exist between 14844, and NUREG-1465.
23 Of particular note is the chemical form of 24 the halogens. Whereas the TID source term was mostly 25 elemental, the NUREG-1465 source term is mostly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 1 aerosols.
2 You can go to slide number now, 12, 3 please.
4 In 2009, after roughly nine years of AST 5 reviews, the NRC proposed changes to Reg Guide 1.183, 6 in the draft Reg Guide known as DG-1199.
7 Multiple changes were proposed, including 8 significant changes to the non-LOCA source term, 9 mostly to address changes in fuel utilization and 10 design.
11 And, that was specifically because of a 12 footnote in Reg Guide 1.183, known as footnote 11, 13 which limited the burn up and the maximal linear heat 14 generation rate, for utilization in the AST 15 applications.
16 The NRC has decided not to finalize DG-17 1199 as Reg Guide 1.183 Revision 1, and is now issuing 18 a replacement known as DG-1389.
19 You can go to slide 13 now, please.
20 DR. BLEY: This is Dennis Bley. Can you 21 give us a little background on why that decision was 22 made?
23 MR. BLUMBERG: Sure. So after the issuance 24 of DT-1199, the staff received a number of public 25 comments, and we spent a significant amount of effort NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 1 in addressing those comments.
2 The efforts included an independent review of 3 certain aspects of DG-1199, that were performed by the 4 Sandia National Lab, report San 2086601.
5 In 2017, the staff received the final 6 responses from SNL, associated with that independent 7 review. But then when we resumed our efforts late in 8 2020, we had gained a significant number of insights 9 since the issuance of that 1199.
10 And, including those 2017 Sandia responses 11 that, where they looked at the San 2086601 report 12 independently.
13 And, then further direction from the 14 Commission in a SECY on back fitting. And, as a also 15 additional research that we had, that had to do with 16 fuel fragmentation, relocation and dispersal.
17 And, so as a result of that, we decided 18 not to finalize DT-1199, and fold all that information 19 into a new set of regulatory guidance, which we call 20 DT-1389.
21 And, that's in process right now. And, in 22 the near future, we'll be coming to the ACRS to 23 provide details on that specific document.
24 DR. BLEY: Thanks, Mark.
25 MR. BLUMBERG: You're welcome.
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23 1 So, as I said just then, DG-1389 builds 2 upon the changes proposed in DG-1199, as modified by 3 public comments.
4 Source term guidance for higher 5 enrichments, which are eight weight percent and burn 6 ups up to 68-gigawatts days per metric ton uranium, 7 peak average burn ups, and near-term accident tolerant 8 fuels, also known as ATFs, are provided in this 9 proposed draft guidance.
10 These near-term ATFs include chromium 11 coated cladding, and chromium dope fuel, but do not 12 include iron chromium aluminum alloy cladding.
13 DG-1389 also addresses fuel fragmentation 14 and relocation, and dispersal.
15 Can you go to slide 14, now, please?
16 MEMBER REMPE: Before you do that, we've 17 not seen the version, the latest version of 1389. But 18 we, anyone can find a copy of draft Guide 1199.
19 And, all of the changes you had on slide 20 13, as well as what I'm seeing here, are LWR based, 21 correct?
22 MR. BLUMBERG: That's correct.
23 MEMBER REMPE: And, so are you planning, I 24 mean I know we've done the pilot planned evaluations 25 and you've got a website.
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24 1 But is staff going to have some guidance 2 on it, and I'm just kind of wondering about if 3 something like what's done for the NPUFs would be, I 4 mean you could give these new design developers some 5 ideas by reflecting on what's done with the NPUFs a 6 bit.
7 MR. BLUMBERG: So, it was my understanding 8 that the ACRS had provide, been provided a draft copy 9 of 1389. Were you, was that made available to you, 10 Joy?
11 MEMBER REMPE: I've not seen it. Dave, 12 have you seen it?
13 CHAIRMAN PETTI: No, I've seen a draft --
14 (Simultaneous speaking.)
15 MEMBER REMPE: I know we're having a 16 meeting in March.
17 CHAIRMAN PETTI: -- of (Audio 18 interference) but I, and I have the draft 1.183 from 19 this past summer when it was earlier on the schedule, 20 and then it got slipped.
21 I don't recall seeing anything on the fuel 22 fragmentation, and how that could have potentially 23 changed the write up. I don't think we've seen that 24 yet.
25 But we're going to be, I'm assuming we're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 1 all going to, that's all going to get wrapped up in 2 the next meeting on 1.183.
3 MR. BLUMBERG: That will be a future --
4 (Simultaneous speaking.)
5 MEMBER REMPE: Yes, well, that's --
6 MR. BLUMBERG: Go ahead.
7 MEMBER REMPE: -- yes, it's coming up in 8 March so we need it soon, is the first question that 9 I had because we do have it scheduled, and you're 10 supposed to get us the documents 30 days in advance.
11 But it's still all LWR based, right?
12 MR. BLUMBERG: It is all LWR based, but it 13 provides ideas on how to deal with future designs.
14 So in other words, it's not giving 15 specific guidance, but it does provide advance 16 guidance for LWR designs that could be potentially 17 used for future designs.
18 MEMBER REMPE: So, it's a bit of stretch 19 for them to come in and try and figure out how to do 20 this, unless they look at.
21 Again, I'm thinking about NPUFs and some 22 other interactions we're having, and I'm just kind of 23 wondering if more could be done. But I guess we'll, 24 that's what we're here to discuss today for the non-25 LWRs, right?
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26 1 MR. BLUMBERG: That's correct. This 2 presentation is an overview on the LWR source term 3 development. And, then future presentations will go 4 into detail about other source terms.
5 MEMBER REMPE: Okay, thank you.
6 MR. BLUMBERG: You're welcome.
7 CHAIRMAN PETTI: Steve, you have a comment?
8 DR. SCHULTZ: Mark, this is Steve Schultz.
9 Question, which may be better in the next 10 presentations to this, to the committee.
11 But you mentioned that some Accident 12 Tolerant Fuel have been incorporated into the draft 13 Guide coming up, and some have not.
14 Why was that, and what is the level of 15 effort that would be required to expand that?
16 MR. BLUMBERG: So, Steve, if you don't 17 mind, if possible I'd like to defer those type of 18 detail discussions to the, to the presentation we're 19 going to give on DG-1389.
20 But in general from a high level, what it 21 boiled down to was lack, lack of data. We don't feel 22 like we've got the data that's necessary to go beyond 23 this 68-gigawatt day for metric ton uranium burn up 24 limit right now.
25 And, also with respect to some of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
27 1 other ATFs, it's the same thing. There's data that 2 needs to be synthesized for other than those near-term 3 designs.
4 DR. SCHULTZ: Good, thanks for the 5 overview. Look forward to the presentation coming up.
6 Thank you.
7 MR. BLUMBERG: Thank you.
8 Okay, so we are on slide 14. And, did we 9 cover 14 already? Let's see here. Yes, and that's 10 basically what slide 14 was about, too.
11 So, due to the limitations on data, 12 currently we're unable to go beyond the 68-gigawatt 13 days per metric ton uranium, and beyond the near-term 14 ATF designs.
15 However, research is under way to 16 accommodate higher burn ups and enrichments, and other 17 ATF designs. And, after that research is completed, 18 our anticipation is that we will come back with a 19 future Reg Guide 1.183 update to include that 20 research.
21 In addition, DG-1389 includes a method for 22 calculating plant specific non-LOCA release fractions.
23 And, includes a generic non-LOCA release fractions for 24 BWRs and PWRs.
25 Please go to slide 15 now.
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28 1 So, to conclude this presentation, I'd 2 like to summarize some of the key points from the 3 presentation.
4 So, one of the ways the NRC staff and 5 licensees determine what measures, and barriers are 6 needed to protect health and safety of the public, is 7 to perform design basis dose calculations.
8 A key component of these analysis is the 9 determination of the release source term. Over time, 10 the NRC has developed regulations, source terms, and 11 regulatory guidance to provide licensees and the 12 staff, with efficient methods of performing these dose 13 analyses.
14 And, then lastly, ongoing efforts by the 15 NRC continue to revise these source terms and methods 16 to address modern fuel utilization, and the use of 17 Accident Tolerant Fuels.
18 So, this concludes my presentation. Are 19 there any additional questions from the ACRS members, 20 that I haven't answered in the presentation?
21 MEMBER KIRCHNER: Mark, this is Walt 22 Kirchner.
23 Just for clarification. So, there's going 24 to be an update to 1.183, as well as a release of DG, 25 let me get the number correct, 1389 in the near NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
29 1 future?
2 MR. BLUMBERG: So, DG-1389 is a draft 3 revision to Reg Guide 1.183, and that's what we'll be 4 presenting to the ACRS for their review in the very 5 near future.
6 MEMBER KIRCHNER: Okay, they're one and the 7 same then?
8 MR. BLUMBERG: They are, they are. And, 9 it's my understanding that you should have access to 10 that draft revision. If you do not, then I'll make 11 sure that you get it.
12 MR. WANG: Hi, this is Weidong Wang, from 13 the ACRS staff. And, we did receive those documents 14 and there was a plan on the November 19 meeting, 15 subcommittee meeting last year. Now it's moved to 16 March 16.
17 So, the files is there.
18 CHAIRMAN PETTI: Okay, so it should be, the 19 question in my mind was yes, I read the ones for the 20 November meeting that then didn't happen in November.
21 It hasn't changed since then?
22 MR. BLUMBERG: Not substantially, no?
23 CHAIRMAN PETTI: Could you check on that?
24 (Audio interference.)
25 CHAIRMAN PETTI: Okay, thanks.
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30 1 MEMBER REMPE: So, what are we supposed to 2 review for March 16? The November document --
3 (Simultaneous speaking.)
4 CHAIRMAN PETTI: Yes. I think so.
5 MEMBER REMPE: -- or, and there won't be 6 anything about the fuel fragmentation, and FFRD stuff?
7 MR. BLUMBERG: Right. The only change that 8 we considered, and we're looking at it right now, was 9 potentially due to the definition of the near-term ATF 10 designs to clarify that. To make sure that it was 11 understood that, that it including the FeCrAl designs.
12 But we haven't actually done that change 13 yet.
14 CHAIRMAN PETTI: Okay.
15 MEMBER REMPE: Okay. And, then --
16 (Simultaneous speaking.)
17 CHAIRMAN PETTI: Because that was an issue 18 that I had raised. And, I thought (Audio 19 interference) like when I heard that, I thought I 20 misread it.
21 But you guys are actually doing a little 22 bit of clarification then, which I think will be good 23 for the committee to hear.
24 MR. BLUMBERG: Good.
25 CHAIRMAN PETTI: So.
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31 1 MEMBER REMPE: And, then --
2 (Simultaneous speaking.)
3 CHAIRMAN PETTI: So, yes, so members, I 4 think that old, that document we got that Weidong 5 talked about is, is fairly current and they'll just, 6 there's just a couple things that they'll probably 7 bring up new in the meeting.
8 MEMBER REMPE: If someone had 9 theoretically, a small modular LWR and they did not 10 want to use the Reg Guide, they wanted to go with a 11 maximum hypothetical accident, does the guidance, this 12 revised guidance, does it help point out what is done?
13 Or needs to be done?
14 MR. BLUMBERG: Could I call on some help 15 here? Michelle Hart, are you on the line?
16 MS. HART: I am.
17 MR. BLUMBERG: Would you like to speak to 18 this?
19 MS. HART: So, I think there is information 20 in Reg Guide 1.183 right now, that talks about the 21 attributes of an acceptable alternative source term.
22 And, so you could use that information to 23 help inform what would be necessary for a maximum 24 hypothetical accident.
25 If you wanted to use the guidance that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 1 out there for non-power reactors, which is more 2 focused on a maximum hypothetical accident as a, as a 3 licensing aspect, you could also use that information.
4 That is NUREG-1537, I think.
5 But there is nothing specific in Reg Guide 6 1.183 that would give you a different licensing source 7 term for that kind of assessment, if that's your 8 question.
9 MEMBER REMPE: That helps. And, I'm just 10 thinking again --
11 (Simultaneous speaking.)
12 CHAIRMAN PETTI: Yes, and --
13 MEMBER REMPE: -- we're supposed to think 14 about things that aren't there yet.
15 I'm wondering if maybe, is it on the 16 website this NUREG-1537 so that, I mean the website's 17 supposed to give everybody all the different 18 references to consider. Is that out there on that 19 website?
20 MS. HART: So, we will talk some more about 21 the website later. I think, you know, right now I 22 can't recall if it's there.
23 But we are intending on keeping that 24 website as up to date as possible, as we figure out 25 new things that need to be on there. Or even old NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 1 things that we, we don't recognize. It's a list of 2 links. So, we'll talk about that some more later.
3 But certainly, it is on the public website 4 now as the standard review plan, and guidance for non-5 power reactors.
6 MEMBER REMPE: But again, we're trying to 7 help the folks that are coming new, and they may not 8 think about it.
9 So anyway, it's just an idea that I was 10 thinking about.
11 MS. HART: Yes, I think this is --
12 (Simultaneous speaking.)
13 CHAIRMAN PETTI: So --
14 MS. HART: -- to event selection kind of 15 discussion.
16 And, so once you figure out the events, 17 you know, you develop your source terms for that. So 18 it's a related topic, or it's an entangled topic we'll 19 say.
20 CHAIRMAN PETTI: So Michelle, in terms of 21 1.183, this guidance, is that the Appendix A as I 22 remember?
23 MS. HART: So, we're --
24 (Simultaneous speaking.)
25 CHAIRMAN PETTI: That talks about the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
34 1 attributes?
2 MS. HART: So, the attributes of an 3 acceptable alternative source term are in position 2, 4 which is in the main body of the text. It's not in an 5 Appendix.
6 CHAIRMAN PETTI: Ahh.
7 MS. HART: The appendices go over each of 8 the accident types, and the source term that would be 9 acceptable. Or, one source term that would be 10 acceptable to meet the regulations.
11 CHAIRMAN PETTI: Okay.
12 So members, I think we're going to hear a 13 lot about LWR stuff, you know, in that meeting here in 14 March. So, I think we should just keep going so we 15 can get through everything.
16 But I see a hand raised. Elijah?
17 MR. DICKSON: Hi, Dr. Rempe. This is 18 Elijah with the staff.
19 We did do some FFRD research in regards to 20 the MHA and the LOCA source term that Mark was 21 discussing.
22 And, in the DG draft guidance that you'll 23 be seeing here shortly, it does refer to some work 24 that the Office of Research had done for us.
25 And, so you can look at that document, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 1 that memo and the analysis that Dr. Mike Solay 2 (phonetic) put together for us, that addresses FFRD 3 for this one particular source term.
4 MEMBER REMPE: So, there's a lot of 5 different draft Guides and things being juggled around 6 here.
7 Again, to make sure we are, because it 8 wasn't in the files that we were given to review for 9 this meeting, and that's why I'm a little puzzled. I 10 didn't know to go back to November.
11 But what we're going to be reviewing in 12 March, will have not only the version of the draft 13 Guide that we're supposed to review, but this 14 additional FFR, I mean I know we've discussed and we 15 actually wrote a letter on FFRD research recently.
16 But it's going to have some additional 17 position or guidance, that will be part of what we're 18 supposed to review, or not?
19 MR. DICKSON: We'll make sure that you have 20 that information, yes.
21 MEMBER REMPE: Okay, thank you.
22 MR. BLUMBERG: And, if you refer to slide 23 9, that's the memo that Elijah was speaking to, the 24 footnote there at the bottom. That provides the 25 reference.
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36 1 It's not slide 9 on your presentation.
2 Just a second, that is slide 13 in your presentation.
3 (Pause.)
4 MEMBER REMPE: So we'll look at that, as 5 well as what we are given for Reg Guide 1.189's 6 update.
7 Thank you.
8 CHAIRMAN PETTI: Yes, so if (Audio 9 interference) that by Dr. Weidong so that we can put 10 it on our SharePoint site, and will all be update.
11 Thank you.
12 Dennis, you had your hand up but then 13 maybe you put it down?
14 DR. BLEY: Right, that's true Dave. I was 15 going to save this for the end, but I'll say it now.
16 This was a nice presentation. It kind of 17 not only showed this history well, but showed the 18 linkages among so many different things people have to 19 tie together.
20 It strikes me, and I'm sure the rest of 21 the presentations will go into the newer material in 22 more detail.
23 Today's transcript and slides might be a 24 really good source for people who have to deal with 25 this issue. Especially some of the new designers.
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37 1 And, I was wondering if among all your 2 other tasks, the staff is thinking of putting together 3 something like a white paper that sort of replicates 4 today's presentations, to give people a good map if 5 they're going to have to do this for themselves.
6 CHAIRMAN PETTI: Yes, Dennis, the word that 7 keeps bouncing in my head is consolidation. Because 8 it's in lots of different places, and lots of 9 different piece parts.
10 And, yes, I worry that the new designers, 11 you know, it's harder to get to that. And, something 12 that would, a road map like you say, could be quite 13 useful.
14 DR. BLEY: Yes, the consolidation is really 15 the key to, well, I think a key for many people on 16 this, to see all the pieces in one place.
17 And, we tracked down a lot of this after 18 we went on what we, we knew from our experience. And, 19 I think it would be very helpful.
20 And, maybe today's meeting slides would 21 serve that same purpose.
22 CHAIRMAN PETTI: I see John Segala has his 23 hand up. John?
24 MR. SEGALA: Yes, thank you.
25 Good comments there. I think that was one NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
38 1 of, that was in our thinking as we were putting 2 together this, this new webpage on source term, was 3 trying to provide, you know, members of the public 4 sort of a here's where all the information is in one 5 place.
6 And, so to the extent that what we do have 7 on the website are our presentations, videos. We have 8 a whole lot of different information up there.
9 So, if today's presentation, slide 10 material, you know, that's all publicly available, I 11 think that's something we could look at adding to, to 12 that website as well.
13 But yes, I think we'll have more 14 discussions as we go through the material today.
15 But thanks for the comment.
16 DR. BLEY: Yes, thanks, John.
17 CHAIRMAN PETTI: Yes, to me --
18 (Simultaneous speaking.)
19 DR. BLEY: I hadn't seen the, have not yet 20 looked at that part of the website, and that might be 21 the perfect place for all of this. And, I look 22 forward to looking at it.
23 CHAIRMAN PETTI: Yes, I mean to me, 24 something like a lead me first, right? The first 25 document you should look at, you know, like the years NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 1 ago when you used to get software, that was the thing 2 you had to do. So that they don't, the staff doesn't 3 get, people don't get overwhelmed, and kind of guides 4 them.
5 Because I think there'll be a lot of 6 information on the website, so, thanks.
7 Okay, let's keep moving along; we're doing 8 good.
9 (Pause.)
10 MR. ESMAILI: Good morning. So, my name is 11 Hossein Esmaili. I am the Chief of Fuel and Source and 12 Code Development Branch, in the Office of Research.
13 I'll be talking briefly about NRC 14 analytical tools and past studies, specifically 15 addressing severe accident progression and source 16 term.
17 Next slide, please.
18 So, before I get into some of the details, 19 I just want to leave you with some high level messages 20 that will be the theme of my presentation.
21 First, we have decades of experience in 22 developing our computational tools that we need to 23 predict the source.
24 These tools are under active development 25 assessment. They are considered state of the practice NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
40 1 tools that is used worldwide, by many regulatory 2 bodies, and research organizations.
3 And, we have a strong relationship with 4 international community to our code sharing programs.
5 This helps us identify knowledge gaps, and participate 6 in experimental research at a fraction of cost to us, 7 to improve our tools.
8 I will go through some of the examples of 9 the more recent supplications of the code, in terms 10 of, especially the work that we have done following 11 the Fukushima accident. And, how these tools have 12 helped us in resolving regulatory issues and 13 decisionmaking.
14 I note that Dr. Petti and his introductory 15 remarks said that there was lot of technical material 16 in this, but my intent is not go into the details of 17 these technical issues, since this have been discussed 18 and presented to ACRS before.
19 But to paint you an overview of the 20 complexities of the issues, and how we go about 21 resolving them using our best state of practice 22 computational tools.
23 Next slide, please.
24 So, first I will give an overview of the 25 code MELCOR. This is our severe accident code. I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 1 will talk briefly about international collaboration, 2 and this is important in how we understand severe 3 accidents, and how we improve our code MELCOR.
4 And, some of the examples as I mentioned 5 in terms of regulatory decisionmaking. This is design 6 certifications for new advanced reactors, state of the 7 art reactor consequence. Now, this is SOARCA, and how 8 it help us you know, move forward.
9 And, some of the post-Fukushima 10 activities. These are the activities that they have 11 been going on for the past 10 years or so.
12 And, finally, I will, Jason will talk 13 about the applications to new advanced reactors. And, 14 this is the work that we have done, the scale MELCOR 15 demonstration calculations.
16 Next slide, please.
17 So, well, next slide.
18 So, the importance of regulatory source 19 term is well established. It finds, I think it's way, 20 into many of our regulations for safety and 21 environmental reviews. Mark was talking about this 22 for LWRs.
23 This slide shows the basics of the source 24 term development process. This is the process that 25 was followed in NUREG-1465,Reg Guide 1.183.
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42 1 And, this is exactly the same process that 2 we are following for advanced nuclear technologies, 3 including ATF and high burn off fuel. And, non-LWRs.
4 So, the first step is the identification 5 of important phenomena, or you know, what will be 6 establish at PERT, phenomena identification, and 7 ranking data.
8 For example, we recently completed a PERT 9 for ATF high burn up fuel, with the aim of 10 understanding knowledge gaps in MELCOR. This is what 11 you see at the, in NUREG 72.83.
12 Of course, we need experimental basis for 13 some of these models in the code. So, we need 14 specific data for fission product diffusivity. And, 15 we model different phenomena.
16 The next step is identification of risk 17 significant accident scenarios. And, this is 18 typically informed by PRAs. NUREG-1150's input to the 19 NUREG-1465, and what accident scenarios we should, we 20 should model.
21 So, what is important in terms of release 22 characteristics is duration of the release, release 23 fraction, and radionuclide species.
24 So, if you are interested in the in 25 containment source term, then we can synthesize the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
43 1 results. This is what we see on the right hand side 2 of the slides.
3 Otherwise, the same process and tools can 4 provide environment release for different accident 5 scenarios. This is typically the NPRAs, or design 6 certifications, or review of chapter 19.
7 Next slide, please.
8 So, this slide shows the relationship 9 between the phenomena we are trying to understand and 10 model, and their experimental basis. This information 11 is then incorporated into MELCOR, by adding and 12 improving modeling capabilities.
13 SCALE is also an important input. It 14 defines decay, hidden ratings like inventories for any 15 MELCOR accident progression analysis.
16 The outputs from the MELCOR is then input 17 into MAACS for offsite consequence.
18 And, I just want to point out a few items 19 here. So, we rely on experiments to understand the 20 phenomena, and validate our models. This is what you 21 see in the dark blue boxes.
22 We have used the codes for regulatory 23 applications. I will go, as I said, I will go through 24 some of the examples in the green boxes.
25 The staff is familiar with the code NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 1 extension to new technologies, including non-LWRs 2 required minimum training, and we have great 3 flexibility for uncertainty analysis.
4 This is becoming more of a standard 5 practice, and this is what we followed in the latest 6 SOARCA analysis.
7 What you see in the red boxes are our 8 focus, non-LWRs and how we're improving our code for 9 non-LWRs and high burn up on Accident Tolerant Fuel.
10 Next slide, please.
11 So, I'm just going to give you a brief 12 overview on history of MELCOR.
13 Next slide, please.
14 So, before MELCOR we had separate effect 15 codes. These where we deemed the source term code 16 package. They were run independently. Results were 17 then manually transferred between the codes.
18 But this led to a number of challenges for 19 transferring data, ensuring consistency in data and 20 properties, and capturing the coupling of physics.
21 For light water reactors over the years, 22 various stand alone codes were integrated into MELCOR.
23 The project actually began in 1982.
24 We had the first release of MELCOR, this 25 is MELCOR 1.80. This was a domestic release in 1986, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
45 1 and then followed by an international release in 1989.
2 In 1991, LANL did a comprehensive peer 3 review was conducted, and all the recommendations are 4 now dealt with. And, code development and maintenance 5 is ongoing, and we typically have annual release of 6 the code as you can see on the slide.
7 At the bottom, I'm just showing you a 8 timeline of the MELCOR development. What models we 9 have put into the code over the past 20 years or so.
10 Some of the milestones include 11 introduction of more mechanistic models. For example, 12 formation of a molten pool in the core. This was 13 introduced in MELCOR 1.86.
14 And, spent fuel pool models. We started 15 looking at spent fuel pools more closely after the 16 9/11.
17 And, new models capturing physics for HDGR 18 started more than 10 years ago in support of NGMP.
19 We also started putting in models for 20 other non-LWR designs, and we were able to conduct 21 public workshops on non-LWR application this past 22 summer, and Jason will talk about those in more 23 details.
24 Okay, next slide, please.
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46 1 level code. It models all phases of a severe accident 2 from accident initiation to core heat up, and 3 degradation leading to vessel and containment failure; 4 tracks fission products through the RCS, and 5 containment to the environment.
6 So, it does everything.
7 Back in the '90s, NRC was developing other 8 codes, such as containment SCDAP. However, these 9 other codes only partially model some of these phases.
10 For example, there are no models for RCS 11 in contained. And, SCDAP it doesn't model containment 12 phenomenon associated with fission product deposition 13 and removal.
14 Research went through a code consolidation 15 phase, and decided to focus on MELCOR. So as you can 16 see on the upper right hand side, all these phenomena 17 are well captured by MELCOR.
18 So, I want to make a very important point 19 here. The question is: what are the requirements for 20 the level of details in a severe accident code?
21 So, when it comes to severe accident, 22 uncertainties in the accident progression and 23 available experimental data for model validation, does 24 not support more detailed model, modeling approach.
25 So, we are able to capture some of this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 1 phenomena through simpler models.
2 The other point is MELCOR is flexible. It 3 allows the user to build the entire model of the plant 4 from basic building blocks, such as control volumes, 5 flow path, and structures.
6 That's why it's easily adoptable to new 7 reactor designs, and implication to spent fuel pools 8 as I showed on the right hand side here.
9 Am I live? Can everybody hear me?
10 CHAIRMAN PETTI: Yes, we can.
11 MR. ESMAILI: I'm not hearing any 12 questions.
13 Okay. Next slide, please.
14 So, again, I have shown a lot of these 15 slides before. So, you're very familiar with it.
16 In this picture, I'm trying to capture the 17 importance in accident progression that can affect the 18 source term. The aim here is to identify similarities 19 and differences within various reactor technologies.
20 So, I also provided a comparison to LWR, and see if we 21 can leverage what we already know, and how we can use 22 our existing tools.
23 I would break it down to three main 24 levels. So, going from the top, when it comes to the 25 containment, there are phenomenon processes that are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
48 1 common. This has to do with high aerosols and vapors 2 evolve and are removed. For example, aerosol-shaped 3 factors may be different under dry and wet conditions 4 in the containment.
5 But when it comes to the primary system, 6 there are, obviously, differences. This is high 7 radionuclides are released from the fuel, but the 8 underlying physics and data is diffusivity to the 9 fuel, and other components can be similar. And we 10 have the coding infrastructure to deal with that. In 11 terms of aerosol dynamics, we also believe similar 12 processes are occurring.
13 There are some ex-vessel phenomena that 14 are different. For example, we have four concrete 15 interactions in the case of LWR and sodium fire in 16 case of SFRs that are unique, that have to be treated 17 separately.
18 Next slide, please.
19 So, code verification and validation is an 20 important element of the software quality assurance.
21 This is the program that we have at Sandia. MELCOR 22 documentation is extensive. There are separate 23 volumes for users' guides, reference or theory manual, 24 and code assessment.
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49 1 physics model, and many of the models that are 2 represented in MELCOR are agnostic to the particular 3 reactor technologies that I showed you before for 4 containment phenomena.
5 Here are a few examples from non-LWR 6 applications. AB1 was the assessment of the codes 7 with AB1 test. This was a test conducted at the 8 Containment System Test Facility in Hanford, 9 Washington. This was a sodium-cooled fire under dry 10 conditions, providing data on aerosol behavior. And 11 as you can see, MELCOR does a reasonable job of 12 predicting the evolution of suspended sodium aerosols 13 over a 50-hour timeframe.
14 We also participated in an IAEA 15 benchmarking looking into releases from the TRISO 16 particles. This is what you see on the upper 17 righthand side of the slide. This is documented in 18 TECDOC-1674.
19 As we move on, we are going to look at the 20 other validations; for example, MSRE in sodium 21 reactors and HTGRs, as we move on.
22 Next slide, please.
23 So, here I'm showing you the evolution of 24 core and RCS nodalizations and the modeling details 25 that we needed over the years.
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50 1 The first input, source input package.
2 This was the model that we had back in the '80s. It 3 has a simple core with one hydrodynamic cell, but 4 multiple core cells in both radial and natural 5 directions.
6 So, when we started in MELCOR, as I 7 mentioned before, in the early '90s, we improved the 8 hydrodynamic modeling for the RCS and the core. By 9 the mid-90s, we had more details on the RCS to model 10 hot leg natural circulation. It became important, you 11 know, when we started looking at steam generator tube 12 rupture. By the end of the '90s, we had more details 13 on the RPBN core to model in-vessel natural 14 circulation, and RCS and in-vessel natural circulation 15 can impact timing of the core damage, hydrogen 16 production, fission product deposition inside the RCS, 17 and potential for RCS piping failure and rupture.
18 So, as we moved through the years, we have 19 updated our practices and our code nodalizations, et 20 cetera. And this is --
21 CHAIRMAN PETTI: Hossein?
22 MR. ESMAILI: Yes, sir.
23 CHAIRMAN PETTI: Just a question. I 24 assume -- I think it's true -- that, as the modeling 25 sophistication has increased, those source terms have, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 1 I guess, reduced. So, the estimates are more 2 accurate, the uncertainty less. Is that a fair 3 assessment generally?
4 MR. ESMAILI: Well, I don't say it's 5 reduced the source term, because I cannot say that 6 because it really depends on a lot of factors. But 7 what I can tell you is that the accident progression 8 affects the source term. So, the more accurate that 9 you model the accident progression -- you know, like 10 the reasons, for example, we have multiple 11 hydrodynamic core cells inside the core was to get a 12 better understanding of how you have oxidation; how 13 that fuel heats up, and how it moves once the core 14 relocation process occurs.
15 So, yes, we have better estimates of the 16 source term.
17 CHAIRMAN PETTI: Yes.
18 MR. ESMAILI: Yes. Does that answer your 19 question?
20 CHAIRMAN PETTI: Yes. I'm just thinking 21 about now you go to these advanced systems, where, 22 hopefully, the physics is simpler; that, you know, the 23 sledgehammer when you may only need to put the tack 24 in. It may be more than is necessarily needed. But 25 MELCOR is scalable in the sense that you could do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 1 something simple, if that's all you felt you needed, 2 right? I mean, it's got that flexibility?
3 MR. ESMAILI: That is true. That is true.
4 And, you know, there's a limit to what we can do. So, 5 when I'm showing this core nodalization, you know, we 6 didn't study it to say that, if you had more models, 7 if you had more nodes, does it improve our modeling.
8 And it turned out that, no, at a certain point, the 9 number of rings, the number of actual levels is 10 sufficient to capture. And this has to do with our 11 inability to know everything about the core relocation 12 process.
13 You are right in terms of some of these 14 non-LWRs, because they don't go through this cliff-15 edge effect, right? I mean, you're not melting.
16 You're not draining this molten. And so, it could be 17 simpler, yes.
18 CHAIRMAN PETTI: Thanks.
19 MR. ESMAILI: Yes, thank you.
20 So, next slide, please.
21 So, for the past few years, we have 22 conducted research to modernize MELCOR code with the 23 key goal of enhancing the efficiency of the 24 development and maintenance of this large, complex 25 code base. Oh, yes, the code is huge. You know, in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
53 1 the earlier presentation, I always refer to it like it 2 has almost 2 million lines of code. It's probably 3 increasing by the day.
4 So, frankly, much of the code architecture 5 reflects programming practices that are several 6 decades old. This does not allow MELCOR to 7 incorporate and benefit from the significant evolution 8 in modern programming languages, operating systems, 9 and compilers.
10 And, in addition, as I mentioned before, 11 the code involved integration of separate other codes.
12 So, these were all integrated into MELCOR, and 13 sometimes using their own solvers. You know, like the 14 other codes that I was mentioning--Maros, VANESSA, et 15 cetera--they had their own solvers because they were 16 standalone codes.
17 But, then, they were integrated into 18 MELCOR in what they call "code packages." The 19 communication between these different packages is also 20 complex, as you can see on the upper righthand side.
21 You know, the hydrodynamics has to communicate with 22 the core package. The core package has to communicate 23 with the radionuclide package, et cetera. So, that 24 communication is pretty sophisticated.
25 So, the idea in modernization, we still NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 1 wanted to have the code as a state of the practice, 2 too. So, the basic idea is to separate the physics 3 from the numerics. You know, numerics is evolving.
4 We don't have to touch all aspects of the code.
5 So, this is done in stages. Last year, we 6 completed the hydrodynamic package. FY22-23, our 7 focus is on the core modeling. You know, this also 8 includes ex-vessel and in-vessel melt progression.
9 And in FY24, we plan to finish with the fission 10 product release and transport modeling, and hopefully, 11 we would have a modernized code by early FY25.
12 All right. Next slide, please.
13 So, this is the slide that I used when I 14 was briefing the ACRS on our, quote, "readiness plan" 15 for non-LWR applications. I'm adding it here. As I 16 mentioned, you know, most of these slides you have 17 seen before, but I'm just putting everything together.
18 So, we have modified the code. We have 19 developed reference models. We have run calculations 20 and conducted workshops. You're going to hear a 21 little bit about that later.
22 So, just like light water reactors, the 23 model that we have developed requires some input data, 24 and the data can come from all sources. It can come 25 from codes, experiments, et cetera. So, I have listed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 1 some of the data needs in terms of what we need to put 2 in the code. But I want to make sure that, as far as 3 the code infrastructure is concerned, it can accept --
4 you know, as data becomes available, we can put it 5 into the code. We do not need to go and change 6 basically the models. We can just change the input 7 parameters.
8 So, as an example of the data needs, the 9 models that are implemented -- but they need data --
10 are fission product diffusivity. Here we have an 11 example of diffusivity through a TRISO particle. We 12 are modeling it; it is simple, one-dimensional 13 diffusion equations. But I need to know what that 14 diffusivity is in terms of calculating the buildup of 15 the fission products through the different layers.
16 SCALE will provide us with the data that 17 we need in terms of generation of the radionuclides 18 into TRISO particles.
19 MEMBER REMPE: Hossein, this is Joy.
20 MR. ESMAILI: Yes, ma'am.
21 MEMBER REMPE: It looks like you wanted a 22 question. So, I'll ask this one now.
23 (Laughter.)
24 I know when we talked about this a while 25 back with the code readiness reviews documents, I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 1 thought we brought up the point that was you do the 2 pilot plant applications. You'll also gain some 3 insights about which data are very important and have 4 more risk impact. And you can wait and answer this 5 later in the pilot plant or reference plant 6 discussions.
7 But I'm curious if you've been able to 8 make much progress in that area, so that it might give 9 some insights to the design developers on, yes, we've 10 got to have something, but plus or minus a factor of 11 two doesn't make much impact on the results. And so, 12 we can use some sort of bounding value for some of the 13 parameter needs.
14 Do you understand where I'm going?
15 MR. ESMAILI: Yes, yes. So, Joy, I don't 16 know whether you have seen the slides. I think in a 17 little bit later slides, I'm talking about some of 18 this uncertainty. And one of the uncertainty analyses 19 that we have done for HTGR, maybe when I get to that, 20 maybe I can answer your question better. Is that --
21 MEMBER REMPE: That's fine, but it's not 22 just that you've identified it. I'm also curious if 23 you're communicating that back. Is there a document 24 on the website that says that -- you know, how do you 25 inform the folks coming in? In this case, you're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
57 1 mentioning the gas reactors. Do the folks that are 2 engaging in the pre-application discussions have a 3 good feel for where the staff is with the progress 4 they've made on the reference plant evaluations?
5 MR. ESMAILI: So, in that, you know we 6 have conducted public workshops, right?
7 MEMBER REMPE: Uh-hum, right.
8 MR. ESMAILI: And we shared information of 9 what we know, you know, what is important. But these 10 are like reference plant models. These are like 11 publicly available information that we can gain.
12 But, as a matter of fact, we are starting 13 to release some of these input models to anybody who 14 wants them. This could be the industry or anybody.
15 And they can follow the same procedure, right, that we 16 have done, to find out what it is that we found 17 important.
18 And so, you know --
19 MEMBER REMPE: You don't have to answer it 20 now --
21 MR. ESMAILI: Okay.
22 MEMBER REMPE: -- but I'm just thinking 23 about readiness and guidance, and some of the things 24 like Dave and Dennis brought up earlier, things that 25 the staff has learned from these evaluations that not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 1 only make them ready for incoming applications, but 2 guidance, so that they focus.
3 You know, if you say that parameters X, Y, 4 and Z are very important -- again, you did a reference 5 plant evaluation -- their design may be different, but 6 they'll know to focus and say, okay, that may be 7 important for what you did, but our design is so 8 different, we don't have to do it. And it might 9 facilitate and make future reviews more efficient.
10 It's just an idea I had. Anyway, just a thought.
11 MR. ESMAILI: I think that is very 12 important, Joy. I really appreciate your asking this 13 question. Now I'm just trying to put this into -- you 14 know, because we are looking at different accident 15 scenarios, you know.
16 But one thing I can say, now that you 17 brought it up, is what I'm showing here in terms of 18 TRISO. For example, we are talking about fission 19 product diffusion coefficients. But the other thing 20 is fuel failure, right? That's another important --
21 so, our experience has shown that we can go ahead and 22 look at these fission product -- and I know that they 23 are doing this under the AGR program with DOE; that 24 they are trying to use their tools to do experiments, 25 et cetera; find out what this fission product NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
59 1 diffusivity is, right? So, this is a large effort to 2 gain that.
3 But it's more important to know what the 4 fuel failure is, right? Because that silicon carbide 5 layer is a good fission product retention, right, 6 layer? So, if you want to know more about this, we 7 would like to know more about the fuel failures. That 8 dominates the source term, as opposed to fission 9 product diffusion coefficients, right?
10 So, these are the insights that we gain 11 from running calculations, doing uncertainty analyses, 12 doing sensitivity. And you're absolutely right, as we 13 go about doing some of these things, we can 14 communicate that, so people or whoever is using our 15 codes understand.
16 Does that --
17 MEMBER REMPE: That helps. It's just a 18 thought to consider, and maybe you've already done it, 19 but something to think about.
20 Thanks. Go ahead.
21 MEMBER BIER: Hi. I have another 22 question, if I can interrupt at this point. This is 23 Vicki Bier.
24 So, I should preface this; I'm not a 25 physical scientist at all. I come from the PRA world.
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60 1 And in the PRA world, there is a certain level of 2 detail beyond which the results really become 3 meaningless. So, like if we tried to expand the level 4 of detail in a PRA to the level of resistors and 5 capacitors in a plant, it would be impossible to 6 create that level of detail of model with any 7 accuracy.
8 And so, as you talk about building more 9 features into MELCOR and more level of detail of what 10 can be modeled, what do you do about validation, and 11 are you creating situations where you are, basically, 12 kind of giving users the ability to hang themselves by 13 giving them so many parameters that may be difficult 14 to estimate or so many models that may be difficult to 15 build accurately?
16 MR. ESMAILI: So, I think, as I mentioned 17 before, you know, we are resisting. We are resisting, 18 and I think of some of the areas mentioned, that we 19 are resisting going into a lot of details. We don't 20 need to do that because, again, MELCOR is a system-21 level code. And, you know, it has been -- it's used 22 consistently for PRA applications. As a matter of 23 fact, we are using it in a Level 3 PRA that we are 24 conducting at NRC.
25 It is -- sorry?
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61 1 MEMBER BIER: No, go ahead. Sorry.
2 MR. ESMAILI: Okay. So, we are very 3 careful in terms of what needs or does not need to be.
4 I will get into that, but there are lots of people all 5 over the world that they use this for their PRA 6 applications. And some people, they go into more 7 details than is necessary. But, through our guidance, 8 we conduct workshops; we meet with them. We are 9 informing them of the level of detail that is needed; 10 the level that you don't really need to get into. And 11 so, we have interaction in terms of what is needed and 12 what is not needed.
13 And we don't ever go into that level of 14 details that you are talking about; that modeling 15 every nut and bolt in the plant. We just model 16 whatever is necessary to gain an understanding of how 17 the accident progresses.
18 MEMBER BIER: Great. Thanks.
19 MR. ESMAILI: Okay. Thank you.
20 Next slide, please, "International 21 Collaboration."
22 So, I'm going to go a little bit -- I 23 don't know how much time I have.
24 But, next slide, please.
25 CHAIRMAN PETTI: Yes, Hossein, we would NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
62 1 probably like to take a break at the top of the hour.
2 MR. ESMAILI: Okay.
3 CHAIRMAN PETTI: So, just to give you a 4 sense.
5 MR. ESMAILI: All right. So, let's see if 6 I can finish by the top of the hour.
7 So, in addition to the domestic use, the 8 code is provided by the NRC to international 9 organizations through bilateral agreements. This is 10 under the Cooperative Severe Accident Research 11 Program, or CSARP.
12 This is an international program. Severe 13 accident, you know, knowledge, code research and code 14 development. It provides access to experimental data 15 for code development, modification, assessment.
16 It's an NRC coordinated program with 17 participation from a lot of countries. Actually, I 18 think that the program started back in the '80s. We 19 have limited experimental programs sponsored by the 20 NRC. The current thrust is on development, 21 assessment, and modification of our tool MELCOR, and 22 we host a meeting once a year, usually in June, to 23 exchange progress in severe accident research and to 24 report code development and assessment status.
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63 1 is that we have approximately about 30 countries that 2 are participating in this program -- you know, from 3 North America to Europe to Asia.
4 There are two user group meetings, one for 5 Asian MELCOR users. It's called Asian MELCOR User 6 Group meeting. This typically meets in the fall, and 7 we had just had one in November. It was hosted in 8 Singapore, and it was still virtual, but hosted by 9 them.
10 And a European MELCOR User Group meeting 11 that meets in the spring. This is sponsored by 12 European countries. So, every year it's different.
13 This year it's going to be Poland who is hosting the 14 meeting. Again, it's going to be virtual and it's 15 going to be sometime in April.
16 These meetings allow more interaction 17 between the code developers and the code users, so 18 they can get more access to what's going on. So, 19 since we are thinking of guidance, when it is time to 20 actually apply the code -- and also, the MELCOR 21 workshop -- this is where we tell people, you know, 22 what it is that they can focus on; what is the 23 important modeling practices; what's the best modeling 24 practices, et cetera.
25 There's also a large user base worldwide.
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64 1 I listed nearly 1,000. So, as you can see from the 2 map, somewhere in the world someone's light is on and 3 probably running one of our codes. So, it has world 4 usage.
5 Next slide, please.
6 So, this slide shows some of the relevant 7 international severe accident projects that the NRC 8 uses to develop and validate the MELCOR code. I'm 9 listing some of them.
10 The PHEBUS, of course, is a very important 11 program involving the knowledge of the tests of 12 fission product release and transport from irradiated 13 reactor vessel, reactor fuel. This was organized by 14 IRSN at the Cadarache facility in France.
15 It consisted of five tests involving 16 releases from irradiated fuel and steam transported to 17 a model RCS, including a steam generator tube and 18 behavior in the model containment. And we use the 19 PHEBUS experiments to validate the MELCOR in 20 NUREG-1465. So, this was one of the first large-scale 21 integral experiments that we were participating in.
22 And from this experimental program, the 23 international severe accident community did a series 24 of separate effects experimental programs to study 25 phenomena for which the code did not capture the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 1 behavior. And we are constantly monitoring, and we 2 are involved in some of these studies.
3 These are the Behavior of Iodine Project, 4 for example; Source Term Evaluation and Mitigation, 5 STEM.
6 And after Fukushima, there were additional 7 programs initiated to better understand the 8 experimental progression with severe accident codes 9 because, as I said in the previous slide, the 10 international community uses our code. So, there's a 11 larger user base of MELCOR, and once you participate, 12 it improves our modeling and predictive capabilities.
13 And recently, we are conducting 14 experiments to study underwater melts. This is 15 especially for composition with high metal content.
16 This is representative of those in Fukushima. This is 17 the ROSAU program.
18 Finally, we are looking into the potential 19 sources of delayed radionuclide releases. This is, 20 you know, where we are talking about revaporization --
21 this was observed at Fukushima -- focusing on the 22 revaporization, the RCS and formation of organic 23 iodides in the containment. This is the ESTER 24 program. They are also participating in an exercise.
25 It's very similar to what we have done under SOARCA.
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66 1 They are looking at uncertainties in severe accidents.
2 Next slide, please.
3 So, last year, we formed a panel of 4 experts. They collaborated to create a Phenomena 5 Identification and Ranking Table. This was to address 6 significant phenomenological issues impacting core 7 degradation and radiological releases for various ATF 8 designs. We also looked at the impact of burnup in 9 enrichment and compared with conventional fuel.
10 The aim of the PIRT was to help the NRC to 11 focus attention on how these concepts change our 12 existing understanding and provide information on how 13 we can use this to improve MELCOR.
14 The final meeting was held, actually, last 15 April, and we published two NUREG documents that you 16 can see on the righthand side of this slide. One was 17 a literature review. This was literature review of 18 what we do know about ATF and high burnup. And the 19 other one was PIRT itself.
20 So, the PIRT really consisted of 10 21 actually internationally recognized experts. So, it's 22 a good document. I highly recommend reading it, and 23 especially the fact that we are looking at these new 24 technologies and comparing it to our conventional 25 fuel.
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67 1 We also worked on calculating the source 2 term for high burnup fuel, and we are planning a peer 3 review for later this year.
4 We are also participating in the 5 QUENCH-ATF program. This is using an experimental 6 facility at the Karlsruhe Institute of Technology, 7 KIT. This is a joint international program through 8 the OECD/NEA. It involves large-scale bunker 9 experiments for design basis and beyond design basis 10 conditions.
11 So, the data that we are getting from 12 these experiments, again, will be used to develop and 13 validate models from MELCOR and FAST. FAST is also 14 our fuel performance code.
15 The project is supported domestically by 16 the NRC, EPRI, and Westinghouse, and internationally 17 by 15 organizations from seven countries.
18 The focus of the first phase would be on 19 chromium-coated zirconium alloys. So, I just learned 20 that the samples have been shipped from Westinghouse 21 to KIT this past January, just about a month ago, and 22 we are hoping to have the first test as soon as April 23 of this year.
24 Next slide, please.
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68 1 program -- this was hosted by IRSN in France -- to 2 investigate loss of coolant or cooling accident in 3 spent fuel pools. This is called the DENOPI project.
4 It's composed of experiments, modeling, validation of 5 computer codes, even CFD codes. These are designed to 6 extend the knowledge about various phases of a loss of 7 cooling accident in the spent fuel pool.
8 The project provides experimental data to 9 validate spray -- I think it's cooling spent fuel, 10 cooling bundles, and cladding oxidation under a 11 mixture of steam and air environments. Specifically, 12 we are interested in rate of spray droplets 13 penetration into the fuel assembly and air and steam 14 oxidation of the cooling cladding to address 15 uncertainty with our current predictive capabilities.
16 We met with IRSN. We looked at some of 17 the results of the experiment, and we are planning to 18 have follow-on to better understand the experimental 19 conditions and how we can improve our code.
20 And this is in line with what we state in 21 SECY-16-0100, following the "National Academy of 22 Sciences Study of the Lessons Learned from the 23 Fukushima Nuclear Accident" for plant users. Also, it 24 is an effort to enhance MELCOR capabilities. So, we 25 are on target with that.
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69 1 Next slide, please.
2 So, I'm going to go look at some of the 3 MELCOR applications.
4 Next slide, please.
5 So, MELCOR is used extensively in the 6 design certification of new reactors. I'm showing you 7 some pictures of what we have used in the past and 8 what are the newer designs. It's used for both severe 9 accident response -- you know, this is mostly when 10 you're talking about the LWRs, Chapter 19, and source 11 term, and the containment response to a design basis 12 accident. This is what we cover usually under Chapter 13 6.
14 The application of MELCOR to these new 15 reactor designs requires some code enhancement and 16 validations because of the special design features.
17 For example -- and this goes back to the '90s -- for 18 example, for AP-1000, we started using a fuel tracking 19 model for the containment shell, because the 20 containment is being cooled on the outside.
21 And as far as severe accident mitigation 22 is concerned, there are design differences between 23 some of these designs. For example, the EPR and APWR.
24 EPR, what you see here, has a special core catcher and 25 spreading compartment, and the containment is equipped NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 1 with passive autocatalytic recombiners for hydrogen 2 control. And the APWR, it's more active systems. The 3 debris shifts through a combination of cavity flooding 4 and enhanced area for debris spreading, and hydrogen 5 control relies on the igniter. So, MELCOR provides us 6 a tool that we can look at these different systems.
7 Next slide, please.
8 This is a good time if you wanted to take 9 a break, because the four slides are SOARCA-related.
10 Or do you want me to keep going?
11 CHAIRMAN PETTI: No, we could probably 12 take a break now. So, thank you.
13 Let's pause here and be back at 15 after 14 the hour.
15 Thank you all.
16 (Whereupon, at 10:56 a.m., the foregoing 17 matter went off the record and went back on the record 18 at 11:15 a.m.)
19 CHAIRMAN PETTI: Okay, Hossein, I have 15 20 after. So, let's keep moving. Thanks.
21 MR. ESMAILI: Okay. Sorry. Dave, do you 22 want me to go a little bit faster in the interest of 23 time? I think I'm a little bit behind, but --
24 MR. SNODDERLY: This is Mike Snodderly.
25 I think you're like four slides behind.
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71 1 I think the goal would be to start the next 2 presentation at noon. But cover the material you need 3 to.
4 MR. ESMAILI: Okay. I'll go a little bit 5 faster because Jason still has to talk.
6 All right. So, back in 2006, we started 7 the State-of-the-Art Reactor Consequence Analysis.
8 This is the SOARCA project. We looked at potential 9 consequences of an unlikely severe accident in Surry 10 and Peach Bottom that involved significant quantities 11 of radioactive material, release of radioactive 12 material into the environment.
13 These two plants are a BWR and a PWR and 14 have been analyzed before in NUREG-1150 and WASH-1400 15 before that. And following completion of the original 16 SOARCA in 2012, we documented the results, and then, 17 we took a systematic look at a potential source of 18 uncertainty.
19 We always acknowledged there are 20 uncertainties in severe accidents, whether in modeling 21 or boundary conditions, and while sensitive analysis 22 is helpful, a formal uncertainty analysis, we thought, 23 could shed more light on the expected behavior.
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72 1 scenarios in Surry, Beach Bottom, and also, for 2 Sequoyah.
3 For the Sequoyah -- this is the picture 4 that you see on the righthand side -- we focused more 5 on the containment performance because of potential 6 for hydrogen combustion and early containment failure.
7 And the picture also shows the details of containment, 8 especially the ice condenser region itself.
9 Next slide, please.
10 So, these are typical results from the 11 SOARCA back in 2012, in terms of source term and 12 release to the environment, as well as containment 13 performance with the ice condenser. The 2000 results 14 show the range of source term for a number of accident 15 scenarios. The UA analysis showed variation in the 16 source term, given uncertainties in the MELCOR 17 modeling frame, as well as boundary conditions, such 18 as safety valve failure for a short-term station 19 blackout in Sequoyah.
20 So, that's what we see in the four steps.
21 There are a few early releases of the iodine due to 22 early failure of the containment, but the majority of 23 the cases resulted in late failure of the containment.
24 So, these are the typical results that you are seeing 25 right now in the uncertainty analysis and detailed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
73 1 modeling with SOARCA.
2 Next slide, please.
3 So, when we were doing the Sequoyah, we 4 started looking more carefully at the UA results and, 5 in particular, trying to find out if we could better 6 understand the response of the system, do a little bit 7 of data mining, and trying to see if we can cluster 8 the points that are important; you know, trying to 9 find out what are the important figures of merit.
10 So, in this case, we were looking at 11 hydrogen generation from the time that the hydrogen 12 started generating after the first deflagration. This 13 is important because it characterizes the accident 14 progression, since only the first hydrogen burn 15 determines early versus late containment failure.
16 And so, on the left, you notice the 17 clustering of the points. So, the yellow are the 18 beginning of cycle for a short-term station blackout, 19 and the red are the long-term station blackouts. So, 20 we were able to cluster these points and superimpose 21 them on top of each other, even though the scenarios 22 were completely different. And this was because, at 23 the time that they started the hydrogen generation, 24 these realizations had similar decay heat and they 25 showed similar behavior.
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74 1 The differences between the middle of 2 cycle and end of cycles are less pronounced. So, the 3 response in terms of hydrogen generation is also 4 similar. So, the point is that, when you look at this 5 clustering, you see that the response is -- you know, 6 the uncertainties are very established. You know, the 7 points are not all over the place and you can find 8 that clustering of the points.
9 During that same Sequoyah, there were also 10 questions of how we modeled the pressurizing safety 11 valve and what impact it has on the results. So, this 12 is what you see on the right figure. And following a 13 briefing of the ACRS on the draft, we revised the 14 model. And so, then, our new model and most recent 15 model showed a probability of 80 percent that the 16 safety valve fails to close. In the revised UA, that 17 probably was reduced to about 11 percent.
18 But both UAs exhibit the same generally 19 characteristics. You know, these are the blue points 20 and the orange points. So, what you see, even though 21 we changed the probabilities, what it did was that it 22 redistributed the points. So, more of the points that 23 are orange now moved into the blue.
24 And this is understandable because the 25 hydrogen that's produced in-vessel is generally more NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
75 1 in the time to hot leg failure and the hydrogen burns 2 longer. Then, the pressurizer safety valve fails to 3 close. And we need a longer time to allow more 4 efficient venting of the hydrogen to the containment.
5 Next slide.
6 So, when we are looking at the Fukushima, 7 we are looking, also, at the clustering of the points.
8 So, we have a true MELCOR that gives us something.
9 I'm not showing initial boundary conditions and event 10 scenario.
11 And some of the questions we can answer 12 are: how well do these different accident management 13 strategies reduce the potential or the magnitude of 14 the release to the environment? What are the critical 15 events in an accident? What are the knowledge gaps in 16 understanding plants' response and various phenomena 17 that have the biggest impact on the accident 18 progression?
19 So, we did this clustering on the 20 righthand side. This is an example of the calculation 21 in the certainty analysis that we did for the AGR that 22 Jason is going to talk in a little bit, and we'll look 23 at some of the, you know, the immersivity of the RCCS, 24 for example; the graphite conductivity; heat transfer 25 from the reactor cavity cooling system, and also, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76 1 blockage; you know, how well we get air cooling of the 2 cavity. And so, we can understand how these 3 parameters affect the results.
4 Next slide, please.
5 So, we also used MELCOR; it played a 6 critical role in forensic evaluation of the three core 7 melt events at Fukushima Daiichi, the events at 8 Fukushima. So, they showed significant uncertainties 9 on our understanding of the boundary conditions.
10 So, what you see here is, with a code like 11 MELCOR, we were able to perform forensic evaluation of 12 the uncertainty boundary conditions and compare 13 available measurements and code estimations to better 14 understand what possible unmeasured and unobserved 15 plant states there were. So, it's very useful in 16 terms of reconstructing the accident that happened at 17 Fukushima. I guess this is the takeaway.
18 Next slide, please.
19 So, we started looking at the spent fuel 20 pool modeling since 9/11, and MELCOR flexibility 21 allowed a relatively straightforward application to 22 spent fuel pool analysis. Over the years, we 23 introduced different components, like, for example, 24 racks and developed a general standard radiation 25 model. And so, this would allow us to better model NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 1 MELCOR for accidents in the spent fuel pools.
2 And like you see, the inserts are some of 3 the NUREG/CRs in terms of zirc fire experiments that 4 we conducted, an international PIRT and status report 5 on the state of modeling for spent fuel pool that I 6 added on the slides.
7 Next slide, please.
8 So, this is an example of the experiment 9 that was done for a PWR assembly. This was an 10 OECD/NEA experiment. This was 12 international 11 partners conducted the experiments.
12 And so, what you can see is that you can 13 see the fuel assemblies. This is the hot channel 14 before the start of the experiments. The hot channel 15 is the first, and it's surrounded by four cooler 16 assemblies. What you can see is that, as it goes 17 through the heatup, the central assembly, you know, it 18 heats up and propagates to the other assemblies, and 19 finally, you can see the state of the assemblies.
20 So, it is also a highly sensitive. So, we 21 tried to model this with MELCOR. Another thing, I 22 think, in general, the MELCOR is capable of predicting 23 the conditions of the experiments. The results were 24 highly sensitive to both oxidation kinetics and 25 transition to breakaway. But, in general, we showed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 1 good agreement as far as peak cladding temperature, 2 ignition time, and ignition propagation is concerned.
3 Next slide, please.
4 So, we did a major study in NUREG-2161 to 5 look at spent fuel pool modeling. Our focus was on a 6 BWR with a Mark 1 containment. We studied the impact 7 of spent fuel configuration, high-density versus low-8 density loading, and we did a very detailed analysis.
9 You can see some of the results here.
10 This is an example of a high-density pool. This is 11 actually at Peach Bottom and it shows where the 12 hottest assemblies from the last two offloads are in 13 a 1x4 configuration and constantly surrounded by 14 colder assemblies.
15 So, we looked at different accident 16 scenarios, a small leak and a moderate leak. When you 17 are like 13 days, the hotter assemblies are hot enough 18 that they can cause serious damage. And as you can 19 see, it heats up and it goes through a zirconium fire, 20 and then, actually, there is fuel relocation.
21 For a moderate leak, this was a case where 22 the break was at the bottom of the pool. So, there 23 was some mitigation associated and natural circulation 24 of air through the fuel assemblies that kept the fuel 25 temperatures relatively lower.
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79 1 Next slide, please.
2 So, in terms of the magnitude of the 3 release, it depends on many factors: rate of 4 draindown, time after out of load, and mode of 5 mitigation. So, you can see this figure shows the 6 environmental release of cesium-137 during that 7 operating cycle. The operating cycle usually for a 8 BWR is about 730 days. And so, we looked at different 9 snapshots during its operating and the range of 10 releases that we can find.
11 In general, they have low-density 12 releases. You know, these are the blue boxes. This 13 is where two-thirds of the assemblies have been 14 removed from the pool. It shows that two orders of 15 magnitude are lower compared to the high-density case, 16 where we had like the fuel completely with the 17 assemblies.
18 What we also found out is that, when we 19 were doing the study, that Peach Bottom is actually 20 not doing 1x4; it's doing 1x8. So, it's one hot 21 assembly surrounded by eight colder assemblies. And 22 this 1x8, it actually was a sensitivity because 23 they're not required to do so, but it was very 24 effective in dissipating the heat through the core 25 assembly. So, this is what we can see in the green NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
80 1 boxes here.
2 But the figure also shows that, during the 3 first week, you know, during the defueling process, 4 the hot assemblies are still pretty hot. So, even 5 though we started doing spray cooling of the 6 assemblies, there was not sufficient water to keep it 7 cool. So, during the first week or so, we still have 8 some releases because we were uniformly spraying 9 throughout at the rate of 200 gpm over the whole pool.
10 Next slide, please.
11 So, during the Fukushima event, it became 12 apparent that reliable vent operation was needed to 13 provide containment integrity. Then, the Commission 14 directed the staff to modify an earlier order on 15 reliable hardened vent to make it be a function of a 16 severe accident. So, we looked at these things 17 carefully under NUREG-2206, "Containment Protection 18 and Release Reduction." We used the model that we 19 developed under SOARCA for Peach Bottom, and we had a 20 run matrix of about 50 runs.
21 We looked at different conditions: the 22 boundary conditions; availability of DC power; how 23 operators control reactor pressure; operation of RCSI 24 from what we learned from the Fukushima accident, and 25 the mitigation during post-core damage. You know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 1 water injection, is it into the driver or the RPV?
2 And how do we vent the containment?
3 So, sometimes you hear the words "severe 4 accident management," "severe accident water 5 addition," "severe accident water management." This 6 is how we do water addition to reduce the releases 7 after lower head failure.
8 Next slide, please.
9 So, in terms of the releases to the 10 environment, we did not explicitly model an external 11 filter, but the effect of filters was considered in 12 the consequence analysis. So here, in the top figure, 13 you see release to the environment. The blue lines 14 are cases without water injection. The red lines 15 assume injections at lower head failure. So, it's 16 clear that, you know, injection is arresting the 17 further releases from the fuel.
18 But what is important also is what you see 19 on the bottom. So, on the bottom, what you see is 20 that, during the release, most of the releases that 21 are occurring, the particles that are being released, 22 it is in submicrons. Like 80 percent at least are in 23 the .5-micron size. Because the release has already 24 gone through the suppression pool, so it becomes more 25 difficult to further scrub.
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82 1 So, this provided some insights into 2 NUREG-2206 and external filtration rulemaking.
3 Next slide, please.
4 So, in terms of the magnitude of the 5 release, these are the range of releases that you 6 find, depending on whether you do RPV injections, 7 driver injections; how you do water exit management or 8 water addition, and if you cycle the vent or not.
9 And as you can see, there's not much 10 different between various post-core damage strategies, 11 whether it's RPV or driver injections, and how you 12 manage the water.
13 One thing is that the highest releases 14 occur for a main steam line break, but this is an 15 unlikely event because they do depressurize the 16 vessel. So, we actually had to work hard; we actually 17 had to induce a steam line break, just to see what the 18 releases looked like.
19 Okay. Next slide, please.
20 So, I just want to close by saying that we 21 have decades of experimental and analytical research 22 in severe accident progression and source term.
23 We know how to develop codes. We 24 understand a lot about how fission products are 25 released and move about and go into the environment, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
83 1 go into the containment.
2 We have a validated state-of-practice 3 MELCOR code. This is ready for application to a wide 4 variety of nuclear technologies, including advanced 5 designs.
6 And it has been an essential tool for 7 resolving safety issues and informing regulatory 8 decisionmaking.
9 Thank you.
10 MR. SCHAPEROW: Okay. I would like to 11 talk to you about MELCOR application in new reactors.
12 Before that, though, I would like to 13 mention that, in case you hadn't noticed, we started 14 MELCOR in 1982. So, this is the 40th anniversary of 15 MELCOR, and MELCOR, I believe, has come a long way.
16 I was there for some of the earlier 17 analyses. The very first MELCOR analysis back in the 18 '90s for steam generator tube rupture, severe 19 accident-induced tube rupture, where we didn't have 20 countercurrent flow in the hot leg, and somebody 21 asked, well, how are the things going to get into the 22 steam generator? How is the steam generator going to 23 heat up? And fission products? So, anyway, this 24 really has come a long way, I believe.
25 So, next slide, please.
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84 1 For new reactors, we do independent 2 analysis using MELCOR. That's part of our review. We 3 have guidance for that. The guidance is given in the 4 Standard Review Plan.
5 The guidance includes: the staff will do 6 an independent assessment of plant response and source 7 term. The staff will do this for select scenarios 8 from the PRA. And then, the staff will sit down with 9 the applicant and discuss any differences or issues 10 the staff might note between the staff's analysis and 11 the applicant's analysis.
12 Next slide.
13 So, we have built a lot of MELCOR models 14 for new reactors. We have MELCOR for all of the large 15 light water reactors. I've listed them here on this 16 slide. I think I've got them all, eight of them. I 17 guess it's maybe more correct to say U.S. EPR and U.S.
18 APWR.
19 The graphic here shows accident 20 progression. This is kind of a graphical depiction.
21 It starts on the upper left and moves around to the 22 bottom left. The first one is the beginning of the 23 accident. The next one is dryout of the steam 24 generator and the core starts to crumble. The one on 25 the bottom right shows the core pretty much all in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 1 lower plenum of the reactor. And then, the one on the 2 lower left shows the core after it's left the reactor 3 vessel and it's sitting at the beginning of the core 4 catcher. And this is the EPR.
5 Next slide.
6 We have also built and applied MELCOR 7 models for SMRs. These are light water reactors.
8 NuScale, mPower, Westinghouse SMR, and more recently 9 and currently, the BWRX-300. We actually use the 10 NuScale model quite a bit, and I'm going to talk about 11 that in the next slide.
12 Next slide, please.
13 So, one thing that we saw with NuScale is 14 that they developed their own source term from the 15 reactor into the containment for the purpose of 16 demonstrating that they meet the EAB and LPZ dose 17 criteria offsite. So, the source term that NuScale 18 came up with was a replacement for the source term in 19 Appendix A of Regulatory Guide 1.183.
20 NuScale did this with MELCOR. They used 21 MELCOR to estimate source term from the reactor into 22 the containment. And part of their analysis and, 23 typically, one of these analyses is the deposition in 24 containment. And for that, actually, NuScale turned 25 from MELCOR and actually decided to use STARNAUA for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
86 1 that. It's a separate code. It's a aerosol code. I 2 understand it was developed years ago in Germany, and 3 they use that.
4 So, when it was our turn to do our 5 independent analysis, we, again, pulled out our tool 6 MELCOR and we performed independent assessment for a 7 number of scenarios from the PRA. And we also used 8 RADTRAD to understand what that meant for offsite 9 doses.
10 Next slide. Okay, next slide, please.
11 So now, I'm going to turn to the MELCOR 12 Scale Non-LWR Source Term Demonstration Project. So, 13 this is where we are in our work for non-LWRs.
14 As Hossein mentioned, we actually started 15 non-LWR work a long time ago. For example, back about 16 10 years ago, we were doing some work on HTGRs. So, 17 here's where we are today:
18 I'm going to talk, starting off with the 19 first bullet here, I'm going to talk about the 20 strategy for non-LWR source term analysis.
21 I'm going to talk about the objectives of 22 this project to develop a MELCOR for non-LWR analysis 23 and to apply it for several reference plant models.
24 I'm going to talk about some public 25 workshops we did this past year.
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87 1 And then, I'm going to show some sample 2 results for the three plants we did this year. I'm 3 only going to show a small sample of the work we did.
4 We did a lot of analyses because we really wanted to 5 give MELCOR a good shakedown, so that we would know 6 that we have a tool that can be used for these 7 reactors.
8 Next slide.
9 So, our strategy for determining source 10 term for non-light water reactors involves applying 11 SCALE in MELCOR. And this is really what we have been 12 doing for many years, pretty much since I've been at 13 the NRC back in the early '90s.
14 In the early '90s, we were using what is 15 called ORIGIN, ORIGIN and MELCOR. And now, it's 16 called SCALE. ORIGIN is part of SCALE. So, we've 17 been using these codes to do source term at least 18 since I got here in the early '90s.
19 Before that, in NUREG-1150, we tended to 20 use -- well, we had a source term code package which 21 was there were separate codes that we had to feed the 22 output of one code into the next code manually.
23 And also, during NUREG-1150, we developed 24 something called the EXOR code package, which, again, 25 wasn't really a -- it was not a phenomenological code NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 1 like MELCOR is. It was factors, you know, release 2 fractions from the core, release deposited in the RCS, 3 et cetera.
4 Anyway, so what we're doing is we're 5 continuing what we've been doing. This is an 6 extension of what we have been doing for many years.
7 We're applying SCALE and MELCOR to the source term.
8 And our strategy for non-LWRs is described in detail 9 in Volume 3 of the code strategy. The cover of that 10 is shown on the left side of this slide.
11 We also plan to use SCALE and MELCOR for 12 safety analysis for fuel cycle facilities for non-13 LWRs. And that's shown on the left side as well.
14 That is so-called Volume 5 of our code strategy.
15 Today, though, I'm going to be focusing on 16 Volume 3 and the work we're doing for the reactors and 17 not for the fuel cycle facilities.
18 Next slide.
19 DR. BLEY: Hey, Jason, this is Dennis 20 Bley.
21 Something both you and Hossein talked 22 about raises a question for me. And that is, over 23 time, as you develop these codes, various people would 24 raise questions about what's there and what's not 25 there, and does this apply to the accident NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
89 1 randomizing. And it caused more evolution as a code 2 in development.
3 Now, for a new design and an applicant 4 who's developed that new design who hasn't been 5 involved in the development of these codes, has access 6 to the website we've heard about today, I'm wondering 7 what keys them to question whether the code's properly 8 handling the physics of anything that's unique in 9 their design. Is there any guidance or help with that 10 that they'll have? Or is it just, when they submit 11 it, you folks might find those things?
12 MR. SCHAPEROW: Yes, well, I think you 13 raise a good point, which is that it is a state-of-14 the-art code. It is a large code. It's 15 sophisticated. It models a lot of phenomena in 16 different levels of detail. And it will be a 17 challenge for a user to code. They need some 18 experience.
19 And various vendors have handled that.
20 One vendor -- I won't name them -- they hired one of 21 the folks at Sandia who's been involved with 22 development in application, when they got started.
23 And I think they trained other people along the way.
24 But, yes, you're right, it is, it will be 25 a challenge for them. But --
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90 1 DR. BLEY: We may run out of those people.
2 MR. SCHAPEROW: Well, that's -- what do 3 they call it these days? -- knowledge management. You 4 know, people are apprenticed, and then, they learn how 5 to do stuff. And then, they develop their skills.
6 Yes, we have MELCOR analysts and 7 developers. At one point, the MELCOR project was 8 really huge. We had people working on MELCOR 9 development at Oak Ridge, Brookhaven, and Sandia. Now 10 Sandia has always been the central spot, and that's 11 where it is now. But, yes, there's a big community.
12 And people that use codes like MAAP, they 13 understand this business quite well, too. They could 14 probably jump right in, I imagine.
15 DR. BLEY: Okay. Well, it's something 16 just for the overview. I mean, this is a session --
17 today's session is on how all of this stuff integrates 18 together, and it's a pretty massive amount of 19 integration.
20 I think we need some kind of guidance that 21 helps people know to look for the kind of things that 22 you folks have talked about that's evolved over the 23 years. Anyway, you don't need to answer that now, but 24 we need to have a focus on that at some point.
25 MR. SCHAPEROW: Well, it looks like NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 1 Hossein would like to add something.
2 MEMBER REMPE: To make it real clear, the 3 best practices volume of the User Guide for MELCOR is 4 something that would be useful for the non-LWR 5 applications, right?
6 MR. SCHAPEROW: Hossein is ready to help 7 out with that one.
8 MR. ESMAILI: So, can I just add something 9 to what you are discussing? So, we have MELCOR 10 workshops, right? So, as long as you are registered 11 code users, you can come to the workshops; you can 12 participate in these codes. You know, this is where 13 Sandia is teaching you how to use the code, if there 14 are additional models for non-LWRs.
15 And this is nothing new. You know, we 16 started having workshops on HTGR and SFR during the 17 pandemic back in the 2018 or 2019 timeframe. So, we 18 always have those training classes for registered 19 users.
20 As I mentioned, we have the theory manual, 21 Users' Guides. You know, the users can go to those to 22 find out exactly what the models are. They can always 23 ask questions. So, they know what the theory behind 24 what we put into the code is.
25 For the surface plant models, we are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 1 working on reports, right? So, those reports are 2 going to become available, right, to see what they 3 have done; you know, when we are using the code, what 4 we found out. And in addition, we can show the 5 models. So, there's a whole bunch of things that are 6 going to come together.
7 But I just want to go back to what Joy 8 said earlier. I think it would be good to put 9 everything in one place in terms of the non-LWRs, to 10 have some type of best practices, et cetera. And I'm 11 sure we can do that.
12 MR. SCHAPEROW: Okay. So, I guess I'd 13 like to turn to the SCALE-MELCOR Non-LWR Demonstration 14 Project. We set up three objectives for this effort.
15 The first is to develop our understanding 16 of severe accident behavior for non-light water 17 reactors, and that would help provide insights for 18 regulatory guidance.
19 Second, we would like to have a dialog 20 with the stakeholders on what we're doing with MELCOR 21 and SCALE and the staff's approach for accident 22 progression and source term. Because we're going to 23 be doing independent analysis, as we always do, I 24 imagine.
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93 1 could use SCALE and MELCOR for non-LWR analysis, and 2 how they can be used to help identify characteristics 3 that are important in an accident, and also, 4 uncertainties that need to be addressed, or maybe that 5 don't need to be addressed.
6 As part of this effort, we have publicly 7 available input models that we're just wrapping up 8 now, at least for three designs. This is new. We've 9 never given plant models out before because the plant 10 models that we have have proprietary information in 11 them.
12 So, these plant models are based on 13 conceptual designs of new reactors that are publicly 14 available, supplemented by things that weren't in the 15 reports. Like the reports, typically, don't specify 16 what building is around the reactor, these design 17 reports for these new reactors. So, we specified a 18 building.
19 Next slide.
20 Regarding the scope of this effort, as I 21 said, we've been doing a lot of work to add new models 22 to MELCOR, such as a heat pipe model. And so, we are 23 developing five full plant models for non-LWRs. We 24 finished the first three this year, this past year.
25 And this current year, 2022, we're doing the last two.
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94 1 This slide shows how we've binned the new 2 reactor designs. We have taken all the new reactor 3 designs and we bin them, roughly, into five classes.
4 And we're building a full plant MELCOR model for each 5 of these five classes. And the individual models that 6 we're building are shown here on this slide.
7 The first one is what was originally 8 called the Megapower reactor that Los Alamos designed, 9 but INL tweaked it and made some changes to it. So, 10 we're modeling the INL Design A version of that.
11 The second reactor, the HTGR, we're 12 modeling the PBMR-400 design, which goes back, again, 13 about 10 years.
14 A newer design is the next one, the 15 University of California Berkeley Mark 1 FHR. This is 16 a pebble-bed reactor with molten salt cooling it.
17 And then, the last two reactors we're 18 going to be doing are the molten salt reactor 19 experiment and the advanced burner test reactor.
20 Next slide.
21 DR. BLEY: Jason, Dennis Bley again.
22 A sort of related question. This is 23 wonderful you're going to have these, and they would 24 be good starting places for many people.
25 Can you talk about -- thinking about the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
95 1 issues of what might be really unique and are we 2 missing anything, more comfort in that area comes with 3 a broader range of peer review and comment. And I 4 don't know the best way to get that. Maybe you can 5 talk about that if you put together an international 6 panel or if you're submitting papers on these that 7 would get international review? Anyway, if you can 8 talk a little bit about the peer review that's going 9 to be applied, or likely will be applied to these 10 models, it would be helpful.
11 MR. SCHAPEROW: Yes. First of all, I 12 think we spent a lot of time developing the code 13 strategy. And the code strategy lays out the models 14 that are needed for these new designs. And so, that's 15 one thing we do factor in.
16 Second is we do have quite a bit of 17 international collaboration going on. Actually, one 18 of the first applications of MELCOR in the non-light 19 water reactor area was by -- I think it was somebody 20 in Hungary. Part of our annual severe accident 21 research meeting, we met and they presented their work 22 on how they applied MELCOR for non-light water 23 reactors. So, yes, there is cross-fertilization, and 24 there is, I guess you might call it, peer review going 25 on there.
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96 1 We've presented the work at public 2 meetings this past year. The vendors came, some of 3 the vendors came. We got some comments. Some of it 4 was more on scenario selection than it was on 5 phenomenological per se.
6 But we're going to be publishing reports, 7 Sandia and Oak Ridge reports, ORNL reports, as part of 8 this project. So, there will be other opportunities.
9 We're presenting results at different conferences.
10 We're producing a new ref conference this 11 year to present one of these results. Our lead 12 analyst, Casey Wagner, is going to be presenting that.
13 So, I don't know if that fully addresses 14 your question, but --
15 DR. BLEY: It's helpful. Thanks a lot.
16 MR. SCHAPEROW: Yes, I think, as with most 17 things, the more we work on it, the more we're going 18 to learn about it and the better our models will get.
19 We're still finding things wrong with the Peace Bottom 20 MELCOR model, little things, you know, at this stage, 21 but there's still little things that get debugged. I 22 guess you might call it bootstrapping. There will be 23 some bootstrapping going on, I think.
24 Anyway, the approach for each of the five 25 designs we're modeling, first, we start off with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 1 SCALE. SCALE is used to calculate what's going on 2 during operation, not during the accident per se, but 3 during operation. So, as the reactor is operating, 4 the gauge is building; radionuclide inventory is 5 building in. So, we're using SCALE to calculate that 6 because that's the starting point for the accident:
7 what's the decay heat? What's the radionuclide 8 inventory in the core?
9 Also, SCALE is providing us reactivity 10 feedback coefficients, which we're using in MELCOR 11 point kinetics models, if we have a reactivity 12 transient.
13 So, the second step is to build a MELCOR 14 full plant model. We're building that based on 15 publicly available designs, design concepts, I should 16 say, that have been proposed over the last number of 17 years.
18 In some cases, we've had to supplement 19 that because these design concept models, they're 20 really are focused on the reactor and not so much on 21 the building. So, running the reactor, and the 22 building surrounding the reactor, of course, is an 23 important place for fission particle deposition.
24 So, the third part is to select accident 25 scenarios. Again, we're turning back to our LWR NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
98 1 experience. We're doing the usual suspects, as you 2 say, the ATWS scenario, the station blackout scenario, 3 and the LOCA with no makeup scenario, those types of 4 scenarios, the, quote, "severe accidents" where 5 everything goes wrong.
6 The fourth step is to perform MELCOR 7 simulations for the selected scenarios, including lots 8 of sensitivity calculations to shake the models down.
9 And also, we've done some uncertainty analysis. We've 10 used MELCOR and we've developed MELCOR input models 11 with -- I shouldn't say that. We varied some of the 12 input parameters with using Monte Carlo tools, and 13 we're running many as MELCOR simulations to look at 14 things with a Monte-Carlo-type approach.
15 And finally, when we finish all that, then 16 we sit down at these public workshops and we discuss 17 what we've done, including the modeling and the 18 results.
19 Next slide.
20 Okay. This is kind of our advertisements 21 for the workshops. We just plopped them here on this 22 slide. The top half shows the workshops we did this 23 year, and then, the bottom half is going to show the 24 workshops we'll do next year.
25 We have a QR code here. If you use your NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
99 1 camera, it plops you right to the website where it is.
2 And also, if you want to see about these workshops 3 some more, there's a link at the bottom.
4 I see a hand raised. Joy?
5 MEMBER REMPE: Yes, in the last slide, I 6 didn't get my hand up in time.
7 MR. SCHAPEROW: Oh, I'm sorry.
8 MEMBER REMPE: And you said, well, we did 9 do sensitivity analyses. And so, I asked the question 10 earlier of Hossein, did you determine which parameters 11 have the most impact? And where there's gaps? And 12 are you communicating that to the folks in the 13 workshop? I listened to some of some of the 14 workshops, and I don't recall that discussion, but 15 maybe I wasn't online at the time it was discussed.
16 MR. SCHAPEROW: No. No. Well, I mean, 17 the first goal of the demonstration calculations was 18 to show that the models we put into MELCOR worked; 19 they work together.
20 MEMBER REMPE: Yes, and I think that's 21 great you guys do that.
22 MR. SCHAPEROW: And, you know, that's a 23 low bar, but MELCOR is a big code and it's a hard job.
24 The second part is to say, well, okay, how 25 do you use the code to look and see maybe what's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
100 1 important, what you really -- what are the results 2 sensitive to? And so, we did demonstrate, I think, 3 how you would do that; how you would use MELCOR to 4 show what the results are sensitive to.
5 But the next step is to show, okay, for 6 this design, what are the results sensitive to? We 7 did a little bit of that, but I think that's kind of 8 the next step. I wouldn't characterize that we've 9 done much of that at this point.
10 And one reason is the designs aren't 11 finished. We had to like make up things. Maybe I 12 shouldn't use the word "make up."
13 MEMBER REMPE: Sure.
14 MR. SCHAPEROW: We had to imagine what the 15 building around the design would like, the building 16 around the reactor. And again, the building can be an 17 important factor in some of these analyses. You know, 18 what happens when the coolant leaves the reactor and 19 enters the building?
20 So, I hope that characterizes it. And 21 maybe, Hossein, I don't know if Hossein had anything 22 to add on this that I would characterize as well.
23 MR. ESMAILI: No, that is beautiful, 24 Jason.
25 MEMBER REMPE: So, again, when I see some NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
101 1 of the applications coming through, I see a lack of 2 data because we all know getting the data is a long 3 lead time and very expensive process. And so, 4 communicating to the design developers what data you 5 had to make up, because they are, also, by the way, 6 making up.
7 MR. SCHAPEROW: Yes, yes.
8 MEMBER REMPE: This input is important, 9 and you'd better have data or show us why you don't 10 have or don't need that data. I think it's a very 11 useful conversation to have with the design developers 12 earlier on, so that we all have as efficient and 13 effective licensing process as possible.
14 MR. SCHAPEROW: And I would suggest that 15 the ACRS actually pointed that out with a recent 16 review of a Source Term Topical Report, that they felt 17 that someone needs to think about that a little more.
18 I think it was release of fission particles from 19 molten salt. So, yes, point well-taken.
20 MEMBER REMPE: I think we're trying to 21 point that out, but it might as well be on this 22 website, or whatever, a roadmap that you guys are 23 developing early on, just to avoid problems in the 24 latter end of things. But, anyway, just a thought.
25 MR. SCHAPEROW: No, thank you.
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102 1 Next slide, please.
2 Okay. So now, I have about, gosh, I don't 3 know how much time. Let me do the first one. I have 4 sample results for three plants. So, I'd like to at 5 least go through the first one. If that's enough, 6 then we can skip the other two.
7 So, I would like to point out, though, 8 before I get into that, that this was definitely a 9 collaborative effort between the NRC, Oak Ridge, and 10 Sandia, Oak Ridge being the experts and the developers 11 of SCALE and Sandia being the experts and developers 12 of MELCOR. And Oak Ridge did the development and 13 running for SCALE for this project, and Sandia did the 14 development and running for MELCOR.
15 Okay. Next slide, please.
16 So, this is the first of the three designs 17 we did this past year. This is the INL Design A of 18 the Megapower reactor. And what I'm showing on this 19 slide is our models for this reactor. The left side 20 of the slide shows the SCALE model which encompasses 21 the reactor. And on the right, the other two graphics 22 are the MELCOR model. All the way on the right is the 23 MELCOR model of the reactor, and it even shows the 24 reflector. This is a fast reactor, so it's got a 25 reflector, and the external B4C shield.
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103 1 And the center of the slide shows the 2 MELCOR nodalization of the reactor building, which, 3 again, we had to kind of make this one up. We decided 4 that we would use something that looks a lot like a 5 BWR Mark 1 reactor building, not the containment, but 6 the reactor building, which has a specified kind of 7 leak rate type of thing.
8 Okay. Next slide.
9 So, we did a few different accident 10 scenarios with MELCOR for the heat pipe reactor 11 design. The one I'm going to talk about today 12 involves a reactivity addition accident with a delayed 13 scram. So, those are the two basic assumptions, 14 scenario assumptions, I would say: that (a) 15 somebody's adding reactivity to this thing 16 unintentionally, and then, the scram doesn't happen 17 right away; it's delayed for about an hour.
18 So, I'd like to direct your attention to 19 the upper righthand graph, the power graph. So, the 20 power of this reactor is 5 megawatts. It's a small, 21 it's a tiny reactor.
22 And so, what happens at the beginning of 23 the accident, we have an inadvertent addition of 24 reactivity by rotating of something called the control 25 drums. So, this is analogous to control rising in an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
104 1 LWR. And so, as the drums turn, reactivity gets added 2 and the power goes up from 5 megawatts up to about 8.
3 And if you look at the bottom graph, 4 you'll see the temperature of the fuel going up, 5 starting around 1100 degrees up to about 1300. And 6 then, there's a knee in the curve at 1300 Kelvin.
7 Because what's going on there is the potassium inside 8 the heat pipes is starting to boil, and the heat pipes 9 aren't working as well now because they're not meant 10 to work with boiling potassium in the pipes. That is 11 kind of the end of them.
12 So, what MELCOR predicts at that point is 13 the temperature really goes up even faster. At 14 2200 K, we have an assumption that the control rods 15 are -- the shutdown rods are inserted. So, that's it.
16 So, the rods go into the core and all that's left is 17 decay heat. And so, after 2200 K, you see the 18 temperature going down because the heat from the fuel 19 is now moving outwards, and the fuel into other 20 materials in the reactor, and then, finally, into the 21 cavity surrounding the reactor. This is a passively 22 cooled reactor. The heat just goes from the outside 23 of the reactor vessel into the cavity, and then, 24 there's air circulation in the cavity.
25 Next slide.
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105 1 So, this is, again, just to give you a 2 flavor of what our MELCOR analysis for a non-light 3 water reactor looks like. We had a three-hour public 4 workshop on this. So, this is really a highlight.
5 Anyway, turning to fission products, this 6 slide talks about fission products. Well, first of 7 all, the top righthand graph shows the pressure in the 8 heat pipes. So, when the heat pipes heat up to 9 2200 K, the pressure goes up from 1 bar to almost 6 10 bar. And then, as the heat pipes fail, then the 11 pressure comes back down to 1 bar.
12 The bottom righthand graph shows what's 13 going on with the iodine. That's, typically, the 14 nuclide of most interest in reactor safety analysis.
15 So, about an hour into the accident, the cladding 16 reaches 1650 K, and that's what we assume at full 17 cladding failure temperature. It's stainless steel.
18 And so, then, we have a release from the fuel, and 19 that's the top curve, the blue curve in the bottom 20 graph.
21 And then, as you can see, the next curve 22 down is the iodine that's in the reactor vessel. So, 23 as you can see, most of the iodine that's released 24 from the fuel stays in the reactor vessel.
25 The next curve down is the green curve.
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106 1 So, that shows how much of the iodine escapes from the 2 reactor vessel into the reactor building.
3 And then, the bottom-most curve is how 4 much escaped into the environment. Now, to have 5 escaping from the vessel to the building, and the 6 building to the environment, you have to have leak 7 paths.
8 So, if you look at the center, the bottom 9 of the center, I show what our assumed leakage was.
10 Again, these are engineering assumptions. We don't 11 know what the tech specs are for the reactor vessel 12 leak rate or the reactor building leak rate. We had 13 to just pick one, so we could demonstrate that MELCOR 14 works and how you might do such an analysis.
15 Next slide.
16 CHAIRMAN PETTI: Jason? Jason?
17 MR. SCHAPEROW: Yes?
18 CHAIRMAN PETTI: Just a question.
19 MR. SCHAPEROW: Sure.
20 CHAIRMAN PETTI: When that heat path from 21 that potassium goes into the secondary system, I don't 22 know if it's an air-cooled system or if it's a water-23 cooled system, but you would know the reaction of the 24 liquid metal?
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107 1 your question, I guess I would like to draw your 2 attention to the center diagram. So, this is our 3 fairly detailed cartoon of our model. The gray area 4 is the fuel. The gray areas are the fuel, and the 5 heat pipe is the green rectangle, the tall, green 6 rectangle.
7 So, for this scenario, we found that the 8 heat pipe failure was in the bottom region. That's 9 where it was hottest. And there was no heat pipe 10 failure up in the top region, where the connection is 11 to the secondary system. So, the release was just 12 into the reactor vessel. And then, the way the 13 fission products made their way into the environment 14 was through the leakage in the reactor vessel wall, 15 and then, the leakage in the reactor building.
16 Now, regarding the question about 17 oxidation of the potassium that left the heat pipe 18 when the heat pipe got a hole in it, I don't believe 19 we modeled it in the simulations. We do have models 20 for metal oxidation in MELCOR. So, if we thought it 21 was an important phenomena, we could certainly 22 activate it or adjust it for potassium. We have one 23 for sodium in there, I'm pretty positive, because we 24 added it a while ago for sodium pool reactors.
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108 1 question.
2 CHAIRMAN PETTI: Yes. No, no, it's, as 3 you talked about the progression of the accident, and 4 this is a really different type of design.
5 MR. SCHAPEROW: Yes, really.
6 CHAIRMAN PETTI: And so, I always worry 7 about that, that material getting out. You know, 8 liquid metal getting out might be the worse thing that 9 happens compared to the fuel.
10 MR. SCHAPEROW: Yes, I don't know.
11 CHAIRMAN PETTI: Yes.
12 MR. SCHAPEROW: The first for me, it was 13 really interesting that we didn't really have what I 14 would call a reactor coolant system anymore. We've 15 got these heat pumps, which kind of is the analog. I 16 should say, instead of one reactor coolant system with 17 one hot leg, you know, now we've got -- I don't know 18 how many of these -- 100 of these or 1,000 of these 19 heat pipes. It's really different.
20 CHAIRMAN PETTI: Yes.
21 MR. SCHAPEROW: So, first of all, we had 22 to add heat pipes to MELCOR. So, we did that, 23 actually, a little over a year ago, a year and a half 24 ago, I guess. Thanks.
25 So, the next two sets of slides I have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
109 1 talk about similar accident analysis for HTGR and for 2 the FHR. Does the Committee want me to go through 3 these? I could. It's just it's going to take more 4 than a few minutes, though. I'm not sure what the 5 Committee would like to hear, since I'm about at the 6 end of my time.
7 CHAIRMAN PETTI: Just hit what you think, 8 if there's certain highlights, you know, that you 9 think are important.
10 MR. SCHAPEROW: Okay. I'll go really fast 11 through them, but I won't go into any detail.
12 So, this is the MELCOR model we put 13 together for a high-temperature gas-cooled reactor 14 and, also, the SCALE model, shown on the left. So, 15 the SCALE model appears on the left. The next one 16 over is the MELCOR model of the reactor, and the one 17 on the upper righthand side, that's the reactor 18 coolant system outside of the reactor. And the bottom 19 diagram is the reactor building.
20 Next slide.
21 So, for this reactor, to demonstrate 22 MELCOR for HTGR, we assume the loss of coolant 23 accident. And that was really it; we assumed the loss 24 of coolant accident.
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110 1 on the top, and then, the bottom viewgraph shows the 2 countercurrent flow in the pipe between the reactor 3 and the break location.
4 Next slide.
5 We assume the control rods went in. But 6 it was a large-break LOCA, so all the heat, all the 7 reactor coolant, all the gases went into the 8 containment.
9 So, the top righthand of the slide shows 10 the flows in the core. The flows were slow initially 11 in the core, but, eventually, they picked up a bit.
12 And this is natural circulation. Because it's a loss 13 of coolant accident, there's no pumping of the gas 14 around the system. The bottom graph shows fuel 15 temperatures predicted by MELCOR at different levels 16 in the core, at different locations.
17 Next slide.
18 Tying to fission products, these are 19 fission product results. Again, the releases are very 20 small. The core didn't get that hot. I mean, these 21 are 10 to the minus 8 iodine releases. I would 22 characterize that as small. The bottom graph is for 23 cesium.
24 Next slide.
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111 1 developed last year was for a pebble-bed reactor with 2 molten salt coolant. The SCALE model is shown on the 3 left. The MELCOR model is shown on the right. So, 4 this shows the core, the reactor coolant system, and 5 the reactor building.
6 Next slide.
7 For this reactor, again, we did a variety 8 of scenarios. We did ATWS. We did station blackout, 9 and we did LOCA. We also did failure of all the -- we 10 pushed this one pretty hard.
11 So, for the ATWS scenario we ran, it was 12 an ATWS. So, the salt pumps shut off, but reactor 13 fails to scram. So now, you're sitting there with an 14 unscrammed reactor.
15 And we also looked at different amounts of 16 decay heat removal from zero up to three full trains.
17 We also included reactivity effects because, again, it 18 was an ATWS and we have what I would characterize as 19 a preliminary analysis, including xenon transient, 20 because xenon is a big deal with you have reactivity 21 accidents. At least that's what we found here.
22 Next slide.
23 On the left side, I show the reactivity 24 predicted in the MELCOR point kinetics model using, 25 again, the SCALE inputs. The first thousand seconds, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
112 1 the reactor is heating up because of the accident and 2 you're getting negative reactivity insertion. And 3 then, after about a thousand seconds, you really see 4 an effect of xenon. The xenon really takes over. And 5 for the next 100,000 seconds, the reactivity is really 6 down because the xenon is adding all kinds of negative 7 reactivity.
8 And so, the next graph over to the right 9 shows the core power going down the first thousand 10 seconds, and then, continuing to go down. At about 11 100,000 seconds -- well, 85,000 to be more precise 12 -- the xenon is kind of gone. And guess what? The 13 reactor's power starts coming up again.
14 So, if you look at the bottom righthand 15 side, you'll see the power going up to up to about 20 16 megawatts-ish, and then, the core heats up and the 17 power comes back down again. And we get oscillatory 18 behavior, which kind of ends up the core power at 19 around 7 or 8 megawatts thermal. That's equal to what 20 the decay heat removal system has taken out. So, 21 that's the new steady state for this reactor, assuming 22 no rods in.
23 Next slide.
24 CHAIRMAN PETTI: Oh, for these sorts of 25 systems, that's exactly what I've seen in the HGTRs, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
113 1 too. You know it's right when the power and the heat 2 removal balance. Then, it's the new steady state 3 until the rods get in, yes.
4 MR. SCHAPEROW: Okay. Next slide.
5 So, this just shows some sensitivities we 6 ran. So, the reactor has a decay heat removal system 7 called DRACS. It's got three trains. So, we ran the 8 accident with all three trains, as I just showed you.
9 And now, we also ran the same accident with two trains 10 working, one train working, and zero trains working.
11 And so, these just show the end trends 12 that you see from that. Maybe one of the more obvious 13 things is on the right side. The fuel temperatures 14 for the case with no trains just keep going up.
15 That's the red line.
16 The initial transient, though, it stops at 17 about 800 degrees C, but, eventually, there's no heat 18 removal. This reactor, without the decay removal 19 system, is kind of in trouble for a severe event like 20 this.
21 Next slide.
22 Summary. We demonstrated the use of SCALE 23 and MELCOR for three classes of non-LWRs. We're 24 working on two more classes this year.
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114 1 what Hossein says, MELCOR simulates the entire course 2 of the accident starting with the initiating event, 3 with the hydraulic response, the fuel heatup, and 4 then, the heat transfer out of the reactor into the 5 reactor building, and then, into the environment. And 6 then, last, but not least, the radiological release.
7 And certainly, MELCOR is unique in this 8 regard. MELCOR has got the necessary aerosol modeling 9 and similar modeling to handle tracking of fission 10 products. It was one of the reasons that the 11 SCDAP/RELAP went away eventually. SCDAP/RELAP did 12 track what left the reactor, what left the core --
13 excuse me -- what fission products, but that was it.
14 After the fission products left the core, they didn't 15 track it. So, we really needed a tool like MELCOR.
16 I'm going back, again, to the '90s.
17 Also, as part of this, we showed that you 18 could use MELCOR to evaluate effectiveness of passive 19 mitigation features, including things such as the 20 DRACS system.
21 That's it.
22 CHAIRMAN PETTI: Hey, Jason, just a 23 comment here. To me, the results also show the 24 importance of the design of the building, right? And 25 there's a message that has to go back to designers, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
115 1 because I think a lot of the advanced reactor 2 designers, you know, everything's passive and I've got 3 this functional containment, and I've got to make it 4 by seismic, but there's an important fission product 5 mitigation capability there that, as they are 6 designing it, they should think about.
7 You guys have to assume something here to 8 just show the calculations, but I think that they 9 could be thinking about this sort of stuff when 10 they're working on those designs. That's important, 11 which they may not appreciate.
12 MR. SCHAPEROW: Also, and going back to 13 maybe the NUREG-1150 days, I understand that the 14 analysis, they included the containment, the 15 deposition and containment, but they didn't always 16 include deposition in surrounding buildings. Because 17 a lot of times, well, the Mark 1, it's got a building 18 around it. So, some of the fission products are going 19 to end up in there. Or Surry, it has an auxiliary 20 building. You know, if there's a pipe break in the 21 auxiliary building, the fission products -- so, even 22 if it's not, quote, "formally credited," or formally 23 whatever, I mean, these buildings exist. And a lot of 24 times, though, they're going to survive the accident.
25 You'll have deposition there.
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116 1 But thank you very much. Anyway, I don't 2 have anything else to present today. So, I guess the 3 next presenter, which I believe is Michelle Hart.
4 CHAIRMAN PETTI: Yes, that's right, Jason.
5 MS. HART: All right. So, yes, my name is 6 Michelle Hart. I work in --
7 CHAIRMAN PETTI: Hold on, Michelle.
8 Just before we start, Arlon, you have the 9 agenda? I don't have it open right now on my 10 computer. When are we supposed to break for lunch?
11 At one o'clock Eastern?
12 MR. COSTA: It's about this time, Dave.
13 MR. SNODDERLY: One o'clock. O'clock 14 Eastern. So, yes, you're 15 minutes behind.
15 CHAIRMAN PETTI: Okay. So, keep going, 16 Michelle.
17 MS. HART: All right. So, I'm Michelle 18 Hart. I work in the Division of Advanced Reactors and 19 Non-Power Production and Utilization Facilities.
20 Next slide.
21 And I'll be talking about the accident 22 source term and near-term in recent applications.
23 So, the outline of this is:
24 First, I'll talk about some recent 25 experience with SMRs and non-light water reactors.
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117 1 Then, we'll talk about the Emergency 2 Planning Zone size justification consequence analyses.
3 I'll give an example of an SMR design 4 certification source term approach.
5 And then, we'll talk about some non-light 6 water reactor early movers.
7 Next slide, please. Slide 82.
8 So, some of our recent experience is that 9 we have been seeing some Topical Report for the SMR 10 for NuScale, for example, and the design certification 11 for NuScale. We've also had some advanced reactor 12 pre-application interactions, which include Topical 13 Report reviews and some license applications, as well 14 as some meetings on special topics. And we have also 15 contracted with the National Labs to develop some 16 Source Term Development Reports.
17 Next slide, please.
18 So, when we talk about the Emergency 19 Planning Zone size justification consequence analyses, 20 the concept is based on NUREG-0396, which is the 21 technical basis for the current regulation which says 22 you should have an Emergency Planning Zone for plume 23 exposure of about 10 miles in radius and an ingestion 24 pathway Emergency Planning Zone of about 50 miles in 25 radius. And that is in the regulation, and it is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
118 1 based on the analysis in NUREG-0396. And it's based 2 on identification of an area within which prompt 3 protective actions may be necessary to provide dose 4 savings in the event of a radiological release.
5 And the feature of the analysis in 6 NUREG-0396 is that it calculates dose at a distance 7 for a spectrum of accidents. These analyses include 8 the design basis accidents and severe accidents.
9 Next slide, please.
10 There were no separate or unique source 11 terms developed especially for EPZ size analysis 12 expected for this. In other words, you kind of re-use 13 the source terms and accident release information that 14 were developed for a safety analysis report or your 15 PRA.
16 Next slide, please.
17 So, we've had some interactions on this 18 with applicants and licensees so far. There's a 19 methodology that TVA developed for the Clinch River 20 ESP Emergency Planning Zone size justification in 21 their Site Safety Analysis Report, and it supports 22 exemptions to the 10-mile plume exposure pathway EPZ 23 side requirement. It did not address ingestion 24 pathway Emergency Planning Zones. And it was purely 25 a methodology. It has not been exercised yet, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
119 1 they have not selected an Emergency Planning Zone for 2 the Clinch River site.
3 We've also had some interactions, and are 4 still undergoing interactions, with NuScale on a 5 Topical Report for EPZ sizing methodology. And that's 6 a methodology to support plume exposure pathway EPZ 7 size determination on a case-by-case basis for 8 reactors under 250 megawatts thermal, which is 9 currently allowed in the regulation.
10 And we also have a rulemaking underway for 11 Emergency Planning Zone size determination for 12 emergency planning for SMRs and other new 13 technologies. And that's in a proposed new regulation 14 50.160. We have just issued, for Commission review 15 and approval, the SECY that goes along with the final 16 rule, and there guidance on analysis in Appendices to 17 the related Regulatory Guide 1.242.
18 I will discuss the NuScale methodology 19 next, and then, I'll talk some more about the 20 rulemaking later.
21 Next slide, please. This is slide 86.
22 So, the NuScale EPZ Sizing Methodology 23 Topical Report, which, as I said, is currently under 24 review, it was not part of the design certification 25 review, but it was submitted at the same time. It's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
120 1 currently on Revision 2. It's applicable to light 2 water reactor SRMs such as NuScale, although it's not 3 limited to the NuScale design.
4 And, in fact, in a response to a recent 5 RAI, the scope of Revision 2 was reduced to advanced 6 light water reactor SMRs. So, it's not for any 7 advanced reactor, as originally envisioned in Revision 8 2.
9 Revision 3 is coming. Some recent 10 discussions focused mostly on consideration of 11 seismically initiated events and the probability of 12 exceeding the acute dose criterion.
13 The applicability of this method to 14 technologies other than light water reactors has still 15 been a topic of discussion with the applicant, but, 16 like I said, they just reduced it to advanced light 17 water reactor SMRs.
18 And the Topical Report is a methodology 19 for the analysis to determine the plume exposure 20 pathway EPZ size. It also does not address ingestion 21 pathway Emergency Planning Zone size.
22 Next slide, please.
23 So, in this Topical Report, source term 24 refers to the fission product release to the 25 environment as a function of time. Unlike with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
121 1 discussions of Reg. Guide 1.183 or TID-14844, in which 2 the source term is released to containment, it 3 includes the containment transport in the 4 determination of the source term.
5 The Topical Report methodology uses source 6 terms from the design basis accidents -- for example, 7 from the NuScale design certification FSAR Chapter 15 8 -- and PRA severe accident scenarios that have been 9 scoped into the analysis according to the methodology.
10 As I had talked about earlier, there's no 11 separate or unique source terms that were developed 12 especially for Emergency Planning Zone size analysis, 13 and the NuScale methodology does use core damage 14 frequency from the PRA to categorize severe accidents 15 and select accident sequences to evaluate against the 16 relevant dose criteria. And so, that's one of the 17 features that makes it maybe not appropriate for non-18 light water reactors which may not have a core damage 19 frequency in their PRA.
20 MEMBER HALNON: Michelle, this is Greg 21 Halnon.
22 MS. HART: Uh-hum.
23 MEMBER HALNON: My understanding is there 24 was a question about the threshold of that CDF and 25 what was below regulatory concern, since there was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
122 1 such a large spectrum of accidents looked at. Was 2 that ever resolved or is that still a question on the 3 table?
4 MS. HART: So, that is still a question on 5 the table. The review is ongoing. And in fact, some 6 of the attempt to resolve some of that issue should be 7 in Revision 3. Unfortunately, I'm not a direct 8 reviewer of that anymore. I'm aware of what's going 9 on, but I can't talk about any details. And because 10 it is still under review, we can't talk about any 11 details.
12 MEMBER HALNON: Okay. Because, during the 13 50.160 discussions, that was one of their sticking 14 points on trying to get the Reg. Guide and some 15 guidance there. We kind of punted towards this 16 Topical Report coming up with some resolution to that 17 issue. So, hopefully, that will be a resolution, but 18 we'll be looking at that in the future, I'm sure.
19 MS. HART: Right. I think there should be 20 a caution, though, about the methodology resolution of 21 the issue may not be a general resolution, because we 22 are taking into account the specific design and the 23 specific design interaction with the environment that 24 may affect that. Because the major concerns that we 25 have are about how to account for external hazards and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
123 1 the probability of those events that are initiated by 2 external hazards. And that may not be a general 3 concern for everyone. And so, there may not be a 4 single way to look at that, but, like I said, it is 5 still something that's still under review.
6 MEMBER HALNON: Okay. Thank you.
7 MS. HART: Okay. So, next slide, please, 8 which should be slide 88.
9 So, a different example -- also for 10 NuScale, though -- is the design certification source 11 term approach that they used. In August of 2019, we 12 did write SECY-19-0079, which describes our review 13 approach to evaluate accident source terms for both 14 the Accident Source Term Topical Report that they had 15 sent in, and we were reviewing at the time, and the 16 design certification application for the NuScale SMR.
17 That SECY talks about how we're going to 18 evaluate their change that they have just recently 19 proposed -- "recently" being at the beginning of 2019 20 -- which moved away from mostly doing what was in Reg.
21 Guide 1.183 with some specific information from the 22 plant to something a little bit more unusual, in that 23 they wanted to use a design basis source term without 24 core damage for environmental qualification of 25 equipment, and for other purposes in the plant. But NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
124 1 they were still using a Reg. Guide 1.183-like source 2 term for the offsite analysis and control room 3 habitability analysis.
4 Next slide, please.
5 So, to talk about the source term 6 methodology, it is an approved methodology for the 7 NuScale SMR design. Its method is to develop accident 8 source terms that are consistent with the Reg. Guide 9 1.183 guidance for PWRs, except for these two special 10 events that they included, which is the core damage 11 source term for the core damage event, which is the 12 equivalent to the maximum hypothetical accident LOCA 13 that we do in Reg. Guide 1.183, Appendix A, and the 14 iodine spike design basis source term, which included 15 no fuel damage, but release of the coolant to the 16 containment.
17 This Topical Report is a reference in the 18 design certification application FSAR, and COL 19 applicants will incorporate by reference the design 20 certification analyses that are listed in the FSAR, if 21 they reference the design certification itself. The 22 Topical Report is not necessarily likely to be 23 implemented separately for COL applications that 24 reference the certified design.
25 The ACRS has reviewed the staff's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
125 1 evaluation of both the Accident Source Term 2 Methodology Topical Report and the accident 3 radiological consequence analyses in the design 4 certification application. Therefore, I'm not going 5 to go into a lot of great detail about it, but I'll 6 summarize some of the important features of the source 7 terms.
8 Staff found that the methods described in 9 the Topical Report to develop design basis accident 10 source terms were acceptable and consistent with the 11 guidance in Reg. Guide 1.183 for PWRs. The two source 12 terms, like I have listed here on the slide -- the 13 core damage source term used to develop the core 14 damage event and the iodine spike design basis source 15 term for use in the evaluation of environmental 16 qualification of equipment -- do not follow specific 17 guidance in Reg. Guide 1.183, but they are generally 18 consistent with the guidance in Reg. Guide 1.183 for 19 similar types of analyses.
20 On the next few slides, I will describe 21 NuScale's methodology to determine the core damage 22 source term and implementation of the design 23 certification FSAR analysis.
24 Next slide, please. So, this should be 25 slide 90.
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126 1 The core damage event was developed to 2 show compliance with 10 CFR 50.47(a)(2)(iv), "The 3 Offsite Dose Criteria" -- you know, the 25-rem that 4 we're all familiar with -- and also address the 5 discussion of fission product release in footnote 3, 6 which is my favorite footnote in the regulation. You 7 know, when you talk about doing this accident analysis 8 to look at offsite consequence analysis, the fission 9 product release should be assumed, for this 10 evaluation, "should be based upon a major accident, 11 hypothesized for purposes of site analysis or 12 postulated from considerations of possible accident 13 events. These accidents have generally been assumed 14 to result in substantial meltdown of the core with 15 subsequent release into the containment of appreciable 16 quantities of fission products."
17 NuScale's core damage event is similar to 18 a maximum hypothetical accident, in that it is not a 19 specific scenario, but is intended to represent an 20 accident with major damage to the core. NuScale's 21 methodology is, in concept, similar to that used to 22 develop NUREG-1465 and the Reg. Guide 1.183 LOCA 23 source term. It is derived from a range of accident 24 scenarios that result in significant damage to the 25 core, informed by the PRA, and it is intended to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
127 1 representative or bounding of a dominant majority of 2 intact containment core damage events for the NuScale 3 nuclear power module.
4 Next slide, please.
5 The core damage event dose analysis in the 6 design certification application FSAR implemented the 7 NuScale Topical Report methodology to determine the 8 core damage source term. And so, it gave the actual 9 release values.
10 The core inventory was calculated using 11 the SCALE code for the core for the NuScale SMR 12 design.
13 And they selected scenarios based on the 14 NuScale SMR PRA, internal events only, and there were 15 five surrogate scenarios that were various failures of 16 ECCS, with decay heat removal system available and an 17 intact containment.
18 Next slide, please.
19 In the NuScale design certification 20 application and in their Topical Report methodology, 21 there was one release phase, unlike in Reg. Guide 22 1.183, which has a gap release phase and an in-vessel 23 phase.
24 MELCOR was used to estimate the release 25 timing and magnitude for each scenario of those five NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
128 1 scenarios, and the release onset and duration from the 2 scenario with the minimum time to core damage was used 3 for the core damage event source term.
4 And the core release fractions were taken 5 from the median of the scenarios for each of the 6 release -- I was going to say, "release categories";
7 that's not the right term -- the species.
8 Time-dependent aerosol removal rates were, 9 then, calculated using the STARNAUA code. They did 10 not use MELCOR for that portion of it. And those are 11 aerosol removal rates within the containment for the 12 SMR.
13 They did use design-specific input from 14 thermal hydraulic conditions that were calculated by 15 MELCOR for a surrogate scenario with the minimum time 16 to core damage.
17 Next slide, please.
18 So now, to move on to source term 19 approaches for non-light water reactor, the ones that 20 we're talking to right now.
21 CHAIRMAN PETTI: Michelle, before we shift 22 there, NuScale kind of lays their own path. But you, 23 the staff, found that it was generally consistent with 24 the methodology in 1.183. I'm struggling with, is the 25 guidance there adequate enough for other SMR vendors?
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129 1 Because there's all sorts of stuff in here they did 2 that is consistent, but is it captured in a functional 3 away? Because there's a lot of design-specific stuff 4 that doesn't make sense. Is there enough there that 5 another SMR vendor could read that and figure their 6 way through to get something that would be acceptable?
7 MS. HART: Right.
8 CHAIRMAN PETTI: So, take what NuScale did 9 and think about it functionally in line with the 10 attributes of a source term, or whatever you want to 11 call it. Is that all there clearly enough, do you 12 think?
13 MS. HART: So, I think, in my opinion, if 14 a light water reactor SMR, or even non-SMR, vendor 15 would look at NuScale's Topical Report methodology, 16 because it is a methodology and it talks about how to 17 go about developing the source terms, that would give 18 them an idea about how to go about developing source 19 terms. And the choices that NuScale made, which is 20 not necessarily true that another SMR vendor may have 21 to make the exact same choices, but it would tell them 22 what they did. And then, they could also look at the 23 information in the design certification application 24 and in our SE to see how it was implemented and the 25 results that came out of that.
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130 1 So, the methodology, when we reviewed the 2 methodology -- and I do admit that we were reviewing 3 the methodology at the same time as we were reviewing 4 its implementation. So, that did help us determine 5 the importance of some of their assumptions.
6 You know, Jason talked earlier about we 7 did use MELCOR and RADTRAD to evaluate some of the 8 questions that we had about the methodology. And we 9 could do that in a more structured kind of way because 10 we had more detailed information on the design itself, 11 because we were reviewing the design at the same time.
12 CHAIRMAN PETTI: Yes, sometimes 13 methodologies can be abstract, right?
14 MS. HART: Yes.
15 CHAIRMAN PETTI: And it's hard to know 16 what it means? Right. I understand that.
17 MEMBER KIRCHNER: Michelle?
18 Okay. Go ahead, Walt.
19 MEMBER KIRCHNER: Yes, thank you, Dave.
20 I thought what was important in the 21 NuScale approach was, consistent with the requirements 22 of 10 CFR 52, they did assume MHA, as you said, that 23 led to core damage. What happens when someone else 24 picking up this methodology does not assume an MHA or 25 assumes that there is no core damage?
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131 1 MS. HART: Right.
2 MEMBER KIRCHNER: Then, how do you 3 proceed?
4 MS. HART: Right. And so, I think, 5 certainly, if somebody is going through development of 6 their plant and development of their licensing 7 strategy, it would really be helpful if they came and 8 talked to us about that and tried to get an early 9 read. You know, they could send in a white paper or 10 they could send in a Topical Report. And even if they 11 wanted to go to the extent of providing us a Topical 12 Report with the source terms results in it, not just 13 a methodology, to get really early "buy-in," quote-14 unquote, on their approach, I think, in general -- and 15 this is something that we're talking amongst ourselves 16 about and kind of developing our position on this --
17 you know, if it's not as obvious that it's a core melt 18 source term with a maximum hypothetical accident, like 19 we're talking about in the footnote, they may have to 20 request an exemption.
21 And depending on the information that they 22 use to justify that, you know, if they have a full 23 structured analysis of the accidents and their 24 facility, and the risk from their facility, they may 25 be able to justify an exemption from that particular NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
132 1 requirement. Because, as we were talking about 2 earlier, you know, there is this idea that, with like 3 the Licensing Modernization Project, that you would do 4 that full assessment, and in conjunction with 5 functional containment, the features of the 6 assumptions behind the regulation, that you would have 7 this large release into a containment, but with a 8 demonstrable leak rate that you would evaluate for 9 this assessment, you know, none of those features may 10 be there for your particular design.
11 Does that help answer the question or? I 12 know it doesn't necessarily give any certainty per se, 13 but it is certainly --
14 MEMBER KIRCHNER: Michelle, there is a 15 statement that what was key, I think, when we looked 16 at this was they did assume an MHA and core damage, 17 and then, proceeded.
18 MS. HART: Right.
19 MEMBER KIRCHNER: The problem I see is 20 that you may have applicants that say, "We don't have 21 a source term as NuScale designs it."
22 MS. HART: Correct.
23 MEMBER KIRCHNER: And then, you're at a 24 juncture where, does the staff have in its evaluation 25 -- like you say, the pre-application meetings between NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
133 1 the applicant and the staff will be very important in 2 resolving it. But I see that as the stumbling block 3 going forward with some of the non-LWR concepts.
4 MS. HART: And I take that as a good 5 point, yes.
6 Are there any more questions about this?
7 CHAIRMAN PETTI: Yes. Because I think 8 this idea of core damage, I mean, could mean just a 9 large fission product release. Because some reactors 10 don't have the same equivalent damage as a light water 11 reactor.
12 MEMBER KIRCHNER: Yes, that is what I 13 mean.
14 CHAIRMAN PETTI: Yes. I mean, think of an 15 HTGR. They're going to come in and they're going to 16 show you really low releases under anything inside the 17 design basis because the fuel is designed to handle 18 all that. And they come in and go, "We don't even 19 have damage, but we are going to postulate something 20 where we get more release." That might, 21 hypothetically, be acceptable because it's kind of 22 meeting the intent of that footnote without the exact 23 detail.
24 MEMBER KIRCHNER: Precisely, Dave, yes.
25 CHAIRMAN PETTI: Right. And I guess the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
134 1 question in my mind is capturing some of that thought 2 process in some paper, the value of doing that for 3 some of these advanced systems. Because sometimes, 4 really, they read it and they go, "Oh, well, that's 5 LWR stuff. We don't need that."
6 But, no, hold it. Step back. Go through 7 that thought process. Figure out what the real intent 8 is behind it, and is the intent behind it well enough 9 documented on the staff side, so that you could have 10 productive discussions going forward? That's sort of 11 a thing in my head.
12 MS. HART: Yes, those are the discussions 13 that we have been having with some of the folks so 14 far. I mean, those exact points are the points that 15 I try to raise and I think we try to raise, that you 16 need to evaluate what potential releases are from your 17 facility and you would use your bounding one.
18 And there may be different ways to go 19 about that. It may be a purely mechanistic analysis.
20 It may be a conservative maximum hypothetical 21 accident, much more like you would see from a non-22 power reactor, where you make unphysical assumptions 23 because you know it would bound anything physical that 24 could happen.
25 I say that guardedly because we have to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
135 1 evaluate what potential things may actually appear.
2 Yes, Dr. Rempe?
3 MEMBER REMPE: Just call me Joy.
4 You've heard several of us mention it 5 would be nice to have more guidance, a roadmap, or 6 whatever. Are you hearing that from the design 7 developers coming in?
8 MS. HART: No, we haven't actually heard 9 that, and I think a lot of them -- I mean, we've only 10 talked to a few. They all have plans to do things 11 like PIRT. They all have plans to do things like 12 evaluate what potential accidents are capable for 13 their reactors. I think they all, the ones that we 14 have coming soon, do plan to talk to us about accident 15 source term and how to go about that.
16 So, I don't know that we're necessarily 17 hearing everything. You know, I can't say that. But 18 we haven't had any specific questions about this, I 19 will say, not to say that maybe they just haven't 20 brought it up because they might not be in that 21 portion of their preparing for licensing yet.
22 MEMBER REMPE: Yes, we're all kind of 23 saying the same thing, but they're not saying it.
24 That's kind of interesting. Thanks.
25 CHAIRMAN PETTI: Yes, I agree with you, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
136 1 Joy. I'm just wondering if it ought to be an item on 2 the Part 53 or to put it on the list to try to get 3 more direct feedback from a broader set of 4 stakeholders. That might be interesting.
5 MEMBER REMPE: But a lot of them are using 6 Part 50 and 52. So, I'm not sure, you know, if Part 7 53 is the only place to think about it.
8 CHAIRMAN PETTI: But whatever is the 9 relevant setting to get that broad stakeholder 10 feedback, I guess that's -- to be more direct and ask 11 that question of them, instead of waiting until you 12 get a pre-application.
13 Okay. Let's keep going then.
14 Oh, hold it. Sorry, I see some hands.
15 John Segala?
16 MR. SEGALA: Yes, this is John Segala from 17 NRR.
18 I was just going to say, you know, we have 19 our periodic advanced reactor stakeholder meetings and 20 we do ask at those meetings if there's need for 21 guidance, and whatnot. So, I think that is something 22 that we could explore at a future meeting. I think 23 the next one is coming up in mid-March.
24 CHAIRMAN PETTI: Okay.
25 MS. HART: All right. So now, to move on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
137 1 to non-light water reactors. The people that we've 2 talked to recently and people we're going to talk to 3 soon -- of course, one of the most recent interactions 4 we've had is with Kairos Power. They did provide us 5 a Topical Report for the mechanistic source term 6 methodology. It's still under review. We did come to 7 the ACRS with a presentation on our evaluation of 8 their Topical Report methodology just last fall.
9 And it's a methodology for applicants to 10 develop event-specific radiological source terms for 11 offsite consequence analysis. It does not include 12 control room habitability or other purposes like 13 normal effluents, or things like that. It's design 14 basis accidents for the siting and safety analysis and 15 AOOs and design basic events for the Licensing 16 Modernization Project process.
17 We are also right now going through a 18 review of the Hermes construction permit application.
19 It evaluates a maximum hypothetical accident, which is 20 a deterministic analysis, and it refers to the Kairos 21 Power Mechanistic Source Term Topical Report for some 22 of the details about the phenomena and how the 23 phenomena are modeled.
24 Kairos, in their construction permit 25 application, did use the SRP for non-power reactors, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
138 1 NUREG-1537, in preparing its application. And the 2 maximum hypothetical accident which they developed is 3 intended to bound all potential accident source terms.
4 I will say, in the construction permit, 5 they are not providing specific consequence analyses 6 or specific highly detailed analyses; that that is 7 something that there is an analysis, but this maximum 8 hypothetical accident may be more bounding than would 9 be expected for an operating license application.
10 It's not refined for the specific design at this 11 point. And like I said, it's under review. We 12 haven't asked initial questions about it yet.
13 Next slide, please.
14 We also have talked to X-energy. They 15 propose to use a developer-made source term code, 16 which they call XSTERM, which includes modeling of 17 radionuclides from generation in the fuel to release 18 and dose.
19 The Topical Report was submitted in April 20 2021, after they had previously submitted a white 21 paper on the concepts that they were talking about for 22 the source term verification and validation for the 23 code suite, which they had submitted back in 2019 and 24 we had given them comments on, or not comments, but 25 feedback. When they submitted their Topical Report in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
139 1 2021, they did address some of those comments.
2 However, we had some questions about the 3 Topical Report, which was to provide the approach 4 taken by X-energy to event-specific mechanistic source 5 terms for their Xe-100 plant. We had questions or 6 discussions with them about the scope of the report, 7 and subsequently, they withdrew the Topical Report in 8 order to revise it and submit a proposed revision to 9 the review scope request. So, we're not currently 10 reviewing anything for X-energy at this time.
11 Next slide, please.
12 As we know, the Oklo Aurora COL 13 application review has recently ended without a 14 resolution at this time. They had proposed a maximum 15 credible accident without release.
16 With respect to the accident source term 17 topic, staff was focusing its review on Oklo's 18 determination of the maximum credible accident 19 scenarios for the Aurora. And so, we had not 20 finalized our evaluation of that at the time we ended 21 our review earlier this year in January.
22 The staff is in pre-application 23 interactions with several other designers, but we do 24 not currently have white papers or Topical Reports on 25 accident source term for review at this time.
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140 1 In a public meeting on January 13th of 2 this year, TerraPower presented a description of their 3 development of source term methodology for the Natrium 4 reactor, and stated they plan to submit a Source Term 5 Topical Report in April of next year. They also 6 stated that they are planning a presentation to the 7 staff on functional containment in April of this year.
8 Source terms and methodologies for other 9 designs have not been submitted to the NRC yet.
10 However, we do have some information on what 11 submittals to expect. For example, the regulatory 12 engagement plan for the Westinghouse eVinci 13 microreactor shows that Westinghouse plans to submit 14 a report on mechanistic or accident source term 15 development that would describe the computer code, the 16 code qualification plan, and outline the methodology 17 that would be used to generate the mechanistic source 18 terms. The submittal timeline is not public.
19 Regulatory engagement plans and other 20 information on pre-application activities, including 21 meetings and submittals on source term, may be found 22 on the NRC's public website under "Nuclear Reactors,"
23 and then, "New Reactors," then "Advanced Reactors,"
24 then "Licensing Activities," then, finally "Pre-25 application Activities". And if you do have a copy of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
141 1 the slides, the blue underlined is a link to that 2 site, and it will tell you what information we have 3 in-house or are expecting.
4 Next slide, please.
5 And that ends my presentation on this 6 particular topic.
7 CHAIRMAN PETTI: Thank you, Michelle.
8 Very helpful.
9 Anybody have any questions before we break 10 for lunch?
11 MEMBER DIMITRIJEVIC: Yes, I do.
12 CHAIRMAN PETTI: Go ahead, Vesna.
13 MEMBER DIMITRIJEVIC: I have a very simple 14 question. What's the relationship between maximum 15 hypothetical accident and maximum credible accident?
16 MS. HART: You know, that's a very good 17 question. Neither term is explicitly defined in the 18 regulations or even in any of our guidance.
19 Maximum hypothetical accident is a term 20 that has been used widely in non-power reactors, in 21 the test reactor world. And in there, if you look in 22 NUREG-1537, it kind of implies that it's an accident 23 that is something greater than would be credible. It 24 may be a non-physical accident. You know, assume that 25 all of the cooling fluid just immediately disappears; NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
142 1 there's no specific reason for it.
2 Whereas, maximum credible accident, which 3 is a term that may have been used in the past by some 4 applicants to refer to like the TID-14844 source term, 5 although some people have called that a maximum 6 hypothetical accident as well, is it's not an accident 7 that would be considered to be exceeded by any 8 accident considered credible. I mean, it should be 9 your most credible or the worst thing that could 10 happen in credible space.
11 But credible is not, and has not been, 12 defined by any specific likelihood or probability of 13 the event or release category, or anything like that, 14 unfortunately. They may or may not be the same 15 accident, but I think there is a flavor to maximum 16 hypothetical that would say that it's something that 17 you think is something that would not occur in the 18 facility.
19 MEMBER DIMITRIJEVIC: So, is it fair to 20 say the maximum hypothetical accident is probably not 21 credible? And what I want to say is, is it fair to 22 say that "credible" is not well-defined in maximum 23 credible, and "maximum" is not well-defined in maximum 24 hypothetical accidents?
25 MS. HART: I think maybe the only thing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
143 1 that may be well-defined is "accident," but, you know, 2 I wouldn't even say that much.
3 (Laughter.)
4 I understand the question, because it is 5 a struggle that we've had, you know, certainly when we 6 were going through the Oklo review, because they were 7 calling their accident "maximum credible." It's a 8 struggle to determine how have they determined that 9 and what do we think are the right things to consider 10 in determining that. And so, I think it's easy to say 11 if something is obviously bounding of anything that 12 could happen in the facility. It's harder to say, 13 when you try to refine that more and get closer to 14 what you think may be an actual expected credible 15 accident, whatever that may mean.
16 MEMBER DIMITRIJEVIC: Thank you. That's 17 a very interesting discussion, especially if you look 18 at these surrogate accidents around NuScale, you know.
19 Are they credible or maximum, or whatever?
20 But thanks.
21 MS. HART: Yes, I think the thing that 22 we're aiming for in that regulation, in that 23 discussion, in that footnote is, you know, let's try 24 to see what would be a bad thing that could happen to 25 test the containment. And if you don't have a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
144 1 specific containment you're testing, well, what would 2 be the bad thing, the worst thing that you would 3 expect to use in your design basis for your siting and 4 for your safety systems that you're crediting in 5 retaining the radionuclides? It's a topic that is 6 very nuanced and may depend on the information that 7 you have in front of you.
8 MEMBER HALNON: So, wouldn't it behoove 9 the staff to try to define those terms, so that it's 10 not across the map in how people use them?
11 MS. HART: I mean, I don't know. I mean, 12 it is something that we are discussing among 13 ourselves. It is something that has come up several 14 times. I don't know that there's a specific 15 definition; like if you use a specific frequency of 16 the event, you know, if that would capture everything 17 that's necessary, or if it's widely acceptable for all 18 potential uses.
19 MEMBER HALNON: Okay.
20 CHAIRMAN PETTI: I see some hands.
21 John? And then, Steve.
22 MR. PARILLO: Yes, this is John Parillo.
23 I just wanted to point out to the 24 Committee that the Draft Guide 1389 does contain a 25 definition of an MHA which is similar to the wording NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
145 1 that Michelle has used, but I can read it.
2 "The maximum hypothetical accident, also 3 referred to as a maximum credible accident, is that 4 accident whose consequences, as measured by the 5 radiation exposure to the surrounding public, would 6 not be exceeded by any other accident whose occurrence 7 during the lifetime of the facility would appear to be 8 credible. As used in this Guide, the term `LOCA' 9 refers to any accident that causes a loss of core 10 cooling. The `MHA LOCA' refers to a loss of core 11 cooling resulting in a substantial meltdown of the 12 core with a subsequent release into containment of 13 appreciable quantities of fission products. These 14 evaluations assume containment integrity with offsite 15 hazards evaluated based on design basis containment 16 leakage."
17 So, that's now in the Draft Guide which we 18 will be presenting, I believe it's March 16th.
19 MEMBER HALNON: Okay. So, the argument is 20 going to be on "credible," not "maximum" or 21 "accident".
22 CHAIRMAN PETTI: But I also think that it 23 depends a little bit on -- that may be a fine 24 definition for water-cooled reactors, but, for 25 advanced reactors, it probably needs some noodling.
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146 1 Because if you have functional containment, it might 2 be a degradation of containment in some way, not 3 necessarily failure, but some of the values somehow 4 get degraded. I mean, there is a lot of, as Michelle 5 says, nuance when you look at it through the lens of 6 an advanced system, I think.
7 MEMBER HALNON: You're going to have to 8 couple some external event to it.
9 CHAIRMAN PETTI: Steve?
10 MR. LYNCH: Sure. Hi. This is Steve 11 Lynch, the Acting Chief of the Advanced Reactor Policy 12 Branch.
13 I just wanted to add a little bit of how 14 we've thought about maximum hypothetical and credible 15 accidents in the non-power world, and how that might 16 be able to be extrapolated into some of the advanced 17 reactor concepts.
18 So, while we don't have an official 19 definition for an MHA or an MCA, we do have a 20 discussion of what to consider when developing one of 21 these accidents in NUREG-1537, both the format and 22 content guide for applicants developing applications, 23 and the Standard Review Plan used by NRC staff to 24 review these applications.
25 And typically, when we're thinking of a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
147 1 hypothetical accident, rather than trying to go down 2 the path of is it credible or non-credible, we've 3 found it helpful to think of it more as, is it a non-4 physical event that is what we are calling the MHA?
5 So, for example, it could be an assumption of a fuel 6 pin exploding in air, where there's not a physical 7 mechanism that would allow for that event to happen.
8 But it does provide a release of fission products that 9 is greater than any design basis accident.
10 And really, when we're looking at MHAs, 11 where that concept is most useful is when you're 12 looking at facilities that have very low consequences 13 to begin with, where you're looking at your suite of 14 design basis accidents, and in the case of non-power 15 facilities, you're not finding a design basis accident 16 that's exceeding 100 millirem at the site boundary.
17 So, then, in order to reduce the burden of some of the 18 analysis that needs to be done, you may look for 19 something that is hypothetical that still 20 demonstrates, even in a non-physical way, that, hey, 21 we're still not exceeding this very low consequence 22 for any accident that could be anticipated at the 23 facility.
24 Then, when we start looking at facilities 25 that might be somewhat larger, and just using examples NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
148 1 of some of the medical radioisotope facilities that 2 are designing to be less than 1 rem at the site 3 boundary, that's when we've started looking at 4 considerations for maximum credible accidents. And 5 we've looked at that as being derived from the design 6 basis accidents at the facility, where there is still 7 an expectation that an applicant has looked at 8 initiating events and all of the accident sequences 9 within families, and then, they select a bounding 10 event from those. And that credible accident is 11 still, more or less, used as a conversation piece and 12 a way of demonstrating the overall risk of the 13 facility, but does not replace any other accident 14 analysis that would otherwise be expected for the 15 facility.
16 So, that's just to give some insights of 17 how the staff has been treating those in some other 18 reviews; that while it hasn't relied on a specific 19 definition, we have had general practices in place for 20 considering these.
21 Thank you.
22 CHAIRMAN PETTI: Joy?
23 MEMBER REMPE: Sure. I'm glad you brought 24 this up, Steve, and I'm glad you went and actually 25 mentioned the fact that sometimes a design basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
149 1 accident that's been evaluated by an applicant might 2 suffice for the maximum hypothetical accident.
3 Because I think there are some good insights, and I'm 4 not sure that those insights are yet on the web page, 5 or whatever guidance the staff's developing. And I am 6 not sure that the design developers that are coming 7 for Part 53 will be looking at the non-power 8 applications, the impact applications. And so, I do 9 think that there's a synergy that could be explored 10 here, especially, again, if they do a systematic 11 approach to come up with the initiating events.
12 But, anyway, I'm glad to hear you mention 13 these insights.
14 CHAIRMAN PETTI: I've captured all this.
15 I think there will be a recommendation coming. Good.
16 Anyone else?
17 (No response.)
18 Okay. Then, let's pause her for lunch and 19 see everyone at 2:00 p.m. Eastern time.
20 Thank you all.
21 (Whereupon, at 1:02 p.m., the foregoing 22 matter went off the record for lunch and went back on 23 the record at 2:00 p.m.)
24 CHAIRMAN PETTI: Okay. It's two o'clock 25 Eastern. Let's continue.
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150 1 Michelle?
2 MS. HART: Hi. Welcome back.
3 So now, for the next topic, we'll be 4 talking about the accident consequence-related 5 regulation activities, and there are several 6 regulation development activities ongoing where source 7 terms will be used. And that's in addition to Part 8 53, which, of course, is happening.
9 Next slide, please.
10 So, the first thing we want to talk about 11 is there is a petition for rulemaking. It was 12 received at the end of 2019 and docketed at the 13 beginning of 2020. And there is The Federal Register 14 notice for it. It's under evaluation. There is no 15 disposition on it yet.
16 The petition requests a voluntary rule to 17 allow power reactor licensees to adopt an alternative 18 to the accident dose criteria specified in 50.67, 19 "Accident Source Term." The petition proposes a 20 uniform value of 100 millisieverts, or 10 rem, for 21 offsite locations and for the control room. So, they 22 would be equal to each other, instead of 25 rem 23 offsite and 5 rem in the control room.
24 Next slide, please.
25 CHAIRMAN PETTI: Michelle, does it say who NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
151 1 petitioned this?
2 MS. HART: So, the petitioner is actually 3 an NRC employee, John Parillo. He is on the line.
4 CHAIRMAN PETTI: Okay. I just wanted to 5 know if it came from industry or where it came from.
6 That's good. Okay.
7 MS. HART: Yes.
8 CHAIRMAN PETTI: Thanks.
9 MS. HART: Yes, there is more detail 10 online. If you go to the rulemaking website, you can 11 see the actual petition that he sent in.
12 CHAIRMAN PETTI: Okay. Thanks.
13 MS. HART: So, the next topic is the 14 emergency preparedness for small modular reactors and 15 other new technologies rulemaking. The stage we're in 16 now is the final rule is in development. And as I had 17 mentioned previously, there is this new section, 18 50.160, and related and conforming changes to other 19 portions of Part 50.
20 The staff presented the rule text and 21 supporting guidance to the Subcommittee in September 22 and the full Committee in November of last year. For 23 the purposes of this topic meeting, I won't be 24 discussing the rulemaking in detail today, but I will 25 describe how the supporting guidance in Reg. Guide NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
152 1 1.242, to be issued with the final rule, currently 2 discusses accident consequence analysis and 3 information on source terms.
4 We did present the draft for Reg. Guide 5 1.242 to the ACRS, which, when issued for public 6 comment, had two appendices: Appendix A, on the 7 methodology for analyses to support determination of 8 plume exposure pathway EPZ size, and Appendix B, which 9 provided information on source terms.
10 Reg. Guide 1.242 includes further 11 clarifications and consideration of the public 12 comments. The changes incorporated into Reg. Guide 13 1.242 were presented to the ACRS last fall.
14 Next slide, please. This should be slide 15 101.
16 Appendix A to the Reg. Guide, "General 17 Methodology for Establishing the Plume Exposure 18 Pathway Emergency Planning Zone Size, provides general 19 guidance on the consequence analysis and discusses 20 event selection and consideration of accident 21 likelihood in determining the scenarios to be used in 22 Emergency Planning Zone size evaluation.
23 In selecting events for the Emergency 24 Planning Zone size evaluation, an applicant should use 25 licensing basis events, such as design basis events, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
153 1 beyond design basis events, and DBAs that they would 2 determine through the LMP process, or the MHA, if they 3 chose to use that in their licensing scheme, and 4 severe accidents, as candidate radiological release 5 scenarios.
6 Considerations include internal and 7 external initiators, all sources of radioactivity 8 release, multi-module and multi-unit considerations, 9 and the event likelihood, including uncertainty, as 10 well as the timing of the releases.
11 Next slide, please.
12 Appendix B to the Reg. Guide, "Development 13 of Information on Source Terms," provides high-level 14 guidance on how to develop source terms for plume 15 exposure pathway EPZ size evaluations. It does not 16 provide specific source terms, but it does give 17 reference to information that could be used to develop 18 source terms. And like we had said previously, the 19 source terms that they would use for these Emergency 20 Planning Zone size determinations would be reuse of 21 source terms they've already analyzed for safety 22 analysis and/or the PRA.
23 Next slide, please.
24 Another rulemaking activity that's ongoing 25 right now is the alternative physical security for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
154 1 advanced reactors. It is a draft rule and guidance in 2 development right now.
3 The proposed rule would establish 4 alternative physical security requirements for 5 advanced reactor technologies that would include SMRs 6 and other non-light water reactors to protect against 7 adversary attack resulting in radiological 8 consequences to public health and safety.
9 So, voluntary alternative physical 10 security requirements would be commensurate with 11 potential consequences to the public health and safety 12 and the common defense and security. The idea is that 13 there would be site-specific analyses demonstrating 14 how the identified alternative physical security 15 requirements meet the applicable performance 16 requirements.
17 This rule has not been issued as a draft 18 rule yet. It's still under development. It should be 19 coming out soon. And so, I can't really describe the 20 details yet, but they will be available for public 21 comment before long.
22 However, I can give you some flavor as 23 to --
24 CHAIRMAN PETTI: Michelle?
25 MS. HART: Yes?
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155 1 CHAIRMAN PETTI: Soon? A month? Six 2 months? What is "soon"?
3 MS. HART: So, it's in the final 4 concurrence kind of status right now. So, I don't 5 have the schedule up in front of me, I'm sorry to say.
6 CHAIRMAN PETTI: Okay.
7 MEMBER BROWN: This is Charlie Brown.
8 What's driving a physical security change 9 for advanced reactors? I mean, a plant is a plant; 10 barbed wire is barbed wire, and fences are fences. I 11 mean, they're still reactors.
12 MS. HART: Right.
13 MEMBER BROWN: That's all I'm curious 14 about.
15 MS. HART: So, you know, the rulemaking 16 plan for this, and the Commission direction on this, 17 was to do a limited-scope rulemaking to address the 18 concerns that some small modular reactors would have 19 in the near term. Before we develop Part 53, there 20 may be a further investigation of how you would 21 include considerations for design in your security for 22 your facility.
23 MEMBER BROWN: I guess I still don't 24 understand. I mean, you're still going to have 25 guards. You've got to have some type of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
156 1 administrative requirements in buildings.
2 MS. HART: So, I am really only here to 3 talk about the source terms.
4 MEMBER BROWN: Okay.
5 MS. HART: I can't really speak to the 6 alternative requirements.
7 MEMBER BROWN: All right. I've got it.
8 MS. HART: That's not what this meeting is 9 about.
10 MEMBER BROWN: Thank you. Okay. I'm 11 done.
12 MEMBER REMPE: Michelle?
13 MS. HART: Yes, Joy?
14 MEMBER REMPE: I'm sorry, maybe you're not 15 allowed to answer the question, but there are Reg.
16 Guides and SRP changes, or is it just Reg. Guides?
17 And can you say which ones?
18 MS. HART: So, we are developing a new 19 Reg. Guide to go along with this rule, and when there 20 are some revisions to a current Reg. Guide on like 21 target set analysis. And so, I don't recall the 22 numbers of the Reg. Guides right off the top of my 23 head, but there are two Draft Guides that would go 24 along with this rulemaking.
25 MEMBER REMPE: Okay. That helps. Thank NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
157 1 you.
2 MS. HART: Uh-hum.
3 To give you a little bit of flavor about 4 the stuff that we're thinking about, though, it is 5 that, you know, kind of similar to --
6 MR. SEGALA: Hey, Michelle?
7 MS. HART: I'm sorry. Yes?
8 MR. SEGALA: Yes, this is John Segala.
9 I just wanted to say, the proposed rule 10 publication date is November 14th, 2022.
11 MS. HART: Thank you, John. I couldn't 12 remember.
13 MR. SEGALA: But it's just now starting to 14 go through the concurrence process as of right now.
15 MS. HART: Yes. Thank you, John.
16 So, with regard to source terms and 17 consequence analysis that may be used -- and, of 18 course, you know, it still hasn't been issued -- but, 19 like with the emergency preparedness rule, we are 20 reusing safety analysis source terms and just 21 determining which scenarios we need to use for that 22 assessment.
23 So, I think you would start with that same 24 basis for the security rule, but there are other 25 considerations you would have to bring into play, such NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
158 1 as potential different release pathways and different 2 initiators. So, that will be something that we would 3 expect to discuss in the Reg. Guides that would go 4 with that rule.
5 Next slide, please.
6 So, that is all the slides I had on this 7 particular topic. Are there any other questions?
8 (No response.)
9 CHAIRMAN PETTI: Hearing none, let's just 10 keep on moving.
11 MS. HART: All right. So, the next topic 12 that we have is guidance and information for 13 developing source terms for non-light water reactors.
14 And in this one, I will have some coworkers that will 15 help with the presentation.
16 Next slide, please.
17 So, the outline of this particular 18 presentation is that we are going to, first, talk 19 about accident consequence analysis for advanced 20 reactors in a general sense.
21 Then, I'll describe mechanistic source 22 term.
23 Next, I'll talk about recent reports on 24 non-light water reactor source term development, those 25 National Lab contractor reports.
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159 1 We'll talk a little bit about the non-2 light water reactor PRA standard and the source term 3 discussion in there.
4 And then, Bill Reckley will talk about the 5 Licensing Modernization Project and the source term 6 that goes along with that.
7 Tim Drzewiecki will talk about an overview 8 of the method in NUREG-2246, "Fuel Qualification for 9 Advanced Reactors," and how that may interact with 10 source term. We'll also discuss that.
11 And then, the last topic, we will talk 12 about this non-light water reactor accident source 13 term information web page that we are developing.
14 Next slide, please.
15 So, when we talk about source terms, 16 including development of mechanistic source terms, 17 we're talking about the source terms for accident 18 assessment. Just as a reminder, there are other 19 source terms and other radiological sources that are 20 part of regulatory assessment, such as those for 21 shielding design or effluents and rad waste system 22 design evaluation, worker protection, things like 23 that. So, when we're talking about source term today, 24 we have only been talking about the accident 25 assessment, and mainly, the offsite consequence NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
160 1 analysis.
2 The siting and safety analysis 3 requirements are the same as used in previous 4 licensing of large light water reactors. We have not 5 changed the regulations for advanced reactors and 6 other new technologies, and the offsite dose 7 consequence reference values and the evaluation of 8 plant design and siting remain the same currently.
9 Newer uses for advanced reactors including 10 the Licensing Modernization Project process for non-11 light water reactors to select licensing basis events; 12 classify system structures and components, and 13 evaluate the adequacy of defense-in-depth as described 14 in NEI 18-04 and endorsed in Reg. Guide 1.233.
15 As we discussed earlier, consequence 16 analysis may be used to aid in plume exposure pathway 17 EPZ size determinations; to support exemptions, or to 18 provide a case-by-case basis for gas-cooled reactors 19 or reactors with rated thermal power less than 250 20 megawatts thermal, in compliance with the current 21 regulations; or, to support applications of the 22 alternative framework for emergency preparedness for 23 SMRs and other new technologies, the 10 CFR 50.160 24 rulemaking, once the rule is issued as final.
25 And as I just discussed, there's an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
161 1 ongoing limited-scope rulemaking for alternative 2 security requirements that use a consequence 3 assessment to say that use of alternative security 4 requirements is acceptable. And the proposed rule and 5 guidance has not been issued yet.
6 Next slide, please.
7 So, some considerations in the development 8 of accident source terms is, of course, you have to 9 determine which events you're going to look at, which 10 scenarios you're going to use for your particular 11 analysis that you're looking for. And I know we've 12 kind of alluded to that a little bit before. Like 13 what scenarios do you use for maximum hypothetical 14 accident? If you're going to use the maximum 15 hypothetical accident, should you analyze radiological 16 consequences for all of your accidents and provide 17 that as your safety analysis? So, the determination 18 there is a part of the basis for determining which 19 source terms you need to develop.
20 When you go through determining which 21 phenomena you need to model, and which systems, 22 structures and components, you would credit in these 23 analyses, there is a balance of prevention versus 24 mitigation of the event. So, certain SSCs are safety-25 related or must be managed in a different way because NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
162 1 they prevent the accidents from occurring. Then, 2 other SSCs are there to reduce the radionuclide 3 release or increase the containment of radionuclides.
4 And so, they have special treatment as well.
5 There's a relationship in determining 6 source terms to the functional containment concept, 7 and the functional containment, as taken from the SECY 8 paper, is a barrier or set of barriers taken together 9 that effectively limit the physical transport of 10 radioactive material to the environment. And that's 11 in place of like a leak-tight physical structure 12 necessarily being required.
13 There's also relationship to PRA, the 14 decisions that you would make with a PRA. You have to 15 develop source terms for some of those decisions.
16 Like if you were to use the non-light water reactor 17 PRA standard, it goes all the way to consequence.
18 You're not making decisions based on proxy metrics 19 such as CDF or LERF. And, of course, in all of this, 20 you need to have a good determination of uncertainty 21 around the physical phenomena and the amount of 22 release and the likelihood of the event.
23 Next slide, please.
24 So, as I had just kind of discussed, 25 there's no requirement that you have a mechanistic or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
163 1 deterministic valuation. That's a choice that a 2 designer can make. If you do use the Licensing 3 Modernization Project, that assumes that you would 4 develop a mechanistic source term and use a PRA.
5 However, there are some non-light water 6 reactors that may choose to provide a postulated 7 maximum hypothetical accident, similar to non-power 8 reactor licensees. The staff acknowledges that 9 there's no specific guidance on development and review 10 of non-light water reactor source terms or mechanistic 11 source terms.
12 However, there is useful information 13 available. Reg. Guide 1.183 has a section, 14 "Attributes of an Acceptable AST," that may be useful, 15 and it describes what attributes are expected for an 16 acceptable alternative source term for light water 17 reactors. Though some aspects of the discussion may 18 not be applicable to some non-light water reactor 19 designs, the guidance is mostly general in nature and 20 should help them be able to address regulatory 21 requirements, if they do look at that.
22 I also note that the guidance in Reg.
23 Guide 1.183 on radiological assessment, in other 24 words, the portions that do not describe the light 25 water reactors source term itself, such as guidance on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
164 1 dose conversion factors, briefing rates, atmospheric 2 dispersion, things like that, is technology-inclusive 3 in manner, and non-light water reactor applicants may 4 find it useful in performance consequence analyses.
5 And we are planning to clarify that in the revision to 6 Reg. Guide 1.183.
7 SECY-93-092, which is a discussion about 8 issues pertaining to advanced reactor designs and the 9 relationship to current regulatory requirements, 10 included staff recommendations on non-light water 11 reactor source terms. I'm going to discuss these 12 recommendations starting on the next slide.
13 If there are any questions?
14 (No response.)
15 If not, next slide, please. It should be 16 slide 110.
17 So, this is the SECY-93-092 definition of 18 mechanistic source term that the staff put forward.
19 "A mechanistic source term is the result of an 20 analysis of fission product release based on the 21 amount of cladding damage, fuel damage, and core 22 damage resulting from specific accident sequences 23 being evaluated. It's developed using best-estimate 24 phenomenological models of the transport of fission 25 products from the fuel to the reactor coolant system NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
165 1 for all holdup volumes and barriers, taking into 2 account mitigation features, and finally, into the 3 environs."
4 Applicants can choose to provide either a 5 mechanistic scenario -- I'm sorry. In the SECY, we 6 recommended that advanced reactor source terms should 7 be based upon a mechanistic analysis, and based on the 8 staff's assurance that certain provisions are met.
9 The SECY defined a mechanistic source term in this 10 manner. As you can see, this definition has a clear 11 relationship to the discussion of functional 12 containment and the barrier assessment that you would 13 do there.
14 Next slide, please.
15 So, the SECY paper also provided these 16 provisions for staff assurance the mechanistic source 17 term was acceptable, or that they were considered.
18 And the first thing is that you should 19 sufficiently well understand the performance of the 20 reactor and fuel under normal and off-normal 21 conditions, so that you can do a mechanistic analysis; 22 and that you should have sufficient data to provide 23 adequate confidence in the mechanistic approach.
24 Transport of fission products can be 25 adequately modeled for all barriers and pathways to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
166 1 the environs; and that the calculation should be as 2 realistic as possible, so that the values and 3 limitations of any mechanism or barrier are not 4 obscured.
5 And that events considered in the analysis 6 to develop the set of source terms for each design are 7 selected to balance severe accidents and design-8 dependent uncertainties.
9 The SECY went on to say that design-10 specific source terms for each accident category would 11 constitute one component for evaluating the 12 acceptability of the design.
13 Next slide, please.
14 So, to go on now to more recent 15 information, we contracted with the National Labs to 16 address accident source terms for non-light water 17 reactors as part of the NRC's non-light water reactor 18 vision and strategy near-term implementation plans, 19 and to respond to the Nuclear Energy Innovation and 20 Modernization Act, or NEIMA. This resulted in these 21 two recent reports that I'll describe now -- one from 22 Idaho National Lab and the other from Sandia.
23 They are technology-inclusive and they 24 tell what to do to develop accident source terms, and 25 not specifically on how to do it. There are no NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
167 1 specific methods or phenomenological models that are 2 presented as "the way to do things." And they do not 3 provide technology-related source terms or releases.
4 In other words, they don't provide specific values 5 that you would expect to see.
6 Next slide, please.
7 The first report is the one from Idaho 8 National Lab. It's titled, "Technology-Inclusive 9 Determination of Mechanistic Source Terms for Offsite 10 Dose-Related Assessments for Advanced Nuclear Reactor 11 Facilities." And it summarizes a risk-informed, 12 performance-based, and technology-inclusive approach 13 to determine source terms. This report was issued in 14 June of 2020, and there is the ADAMS Report Number, 15 Accession Number.
16 It's a graded process that allows both the 17 non-mechanistic source term calculation methods which 18 adopt conservative approaches and assumptions based on 19 known physical and chemical principles, and more 20 importantly, the mechanistic source term calculation 21 methods, which consider design-specific scenarios and 22 use best-estimate models with uncertainty 23 quantification for a range of licensing basis events 24 to be used for the design and licensing of advanced 25 nuclear technologies.
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168 1 Accident source terms are developed to 2 address licensing issues to support the application 3 processes of 10 CFR Part 50 for a construction permit 4 and operating license or 10 CFR Part 52 for combined 5 operating license, standard design certification, 6 early site permit, or standard design approval, or 7 manufacturing license. They can also be used for 8 other purposes, including equipment and environmental 9 qualification, control room habitability analysis, of 10 assessments of severe accident risks and Environmental 11 Impact Statements.
12 Next slide, please.
13 This is taken from page 6 of the INL 14 report. It's a picture of the mechanistic source term 15 as being developed through a systematic evaluation of 16 transport through barriers to the radiological release 17 to the environment. It's a picture that they took, 18 reproduced from a Sandia report, which is the next 19 report I'll be talking about.
20 As you can see, it looks at each of the 21 barriers, and once you figure out what the release is 22 from each barrier and the retention in that barrier, 23 it would result in the release to the environment.
24 Next slide, please.
25 This is another figure that is taken from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
169 1 the INL report. And it shows all of the different 2 considerations and information that you would use to 3 develop source terms to compare to regulatory 4 criteria. And I know it's a nice picture, and it is 5 described in more detail, or the concepts are 6 described in more detail in the paper.
7 Next slide, please.
8 So, the INL report describes some steps in 9 the methodology to develop mechanistic source term.
10 The steps to developing a complete mechanistic source 11 term were developed with the Licensing Modernization 12 Project in mind. So, it includes determination of 13 licensing basis events, as the term is used in 14 NEI 18-04.
15 The steps allow for flexibility to 16 accommodate refinements in the source term 17 determination, as needed for the specific purpose, 18 including allowing for non-mechanistic or simplified 19 mechanistic approach.
20 One approach is to use the initial 21 bounding calculations from step 4 to meet 22 requirements. If they meet requirements that are 23 sufficient for your licensing purpose, that should be 24 sufficient.
25 This is intended for facilities that have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
170 1 a small enough initial inventory of source terms to 2 meet radiological requirements upon a full release of 3 the initial inventory. So, we don't expect it's 4 something that the majority of designs may be able to 5 use.
6 The second pathway can use the System 7 Hazard Analysis, or SHA, and perform simplified 8 calculations, as noted in step 5, to identify barriers 9 and a maximum fractional release to perform a 10 simplified mechanistic bounding analysis that could 11 meet radiological control requirements -- or 12 radiological consequence requirements. Excuse me.
13 A third pathway, which is still not a full 14 mechanistic source term approach, is to use the loop 15 of redesign in step 6, after going through step 5, and 16 following through back again to step 4, and then, back 17 to step 5, to continue to refine your analysis. If 18 these pathways are not sufficient, a complete 19 mechanistic source term approach going all the way 20 through to, well, 14 -- you should always document 21 your source term development -- would be included.
22 So, as you can see from these steps, there 23 is a set of systematic evaluations that you go through 24 and you refine and you make it more detailed, and 25 consider the uncertainty and the design goals, as you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
171 1 go through this assessment.
2 And I know I went through that really 3 fast, and I don't know that it necessarily made sense.
4 So, do you have any questions?
5 (No response.)
6 If not, we'll go to the next slide.
7 Now I will talk about the lab report that 8 we've gotten, which is a "Simplified Approach for 9 Scoping Assessment of Non-Light Water Reactor Source 10 Terms." And this was issued by Sandia in January of 11 2020. And that's the ADAMS Accession Number for that.
12 This approach is intended to identify the 13 characteristics of reactor design concepts, release 14 mitigation strategies that are most important to 15 different classes of accident scenarios. And it uses 16 a scoping methodology to provide an approximate, 17 order-of-magnitude estimate of the radiological 18 release to the environment and associated offsite 19 consequences.
20 The scoping method is applied to different 21 reactor concepts, considering the performance of 22 barriers to fission product release for these concepts 23 under sample accident scenarios. The accident 24 scenarios and sensitivity evaluations are selected in 25 this report to evaluate the role of different fission NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
172 1 product barriers in reducing the source term to the 2 environment and associated offsite consequences.
3 It did not develop quantitative estimates 4 of radiological release magnitudes and compositions, 5 and I did say it looked at several different concepts.
6 Those concepts were high-temperature gas reactors, 7 sodium fast reactors, and liquid-fueled molten salt 8 reactors. And it is primarily qualitative.
9 Do you have any questions about this 10 report?
11 (No response.)
12 Okay. The next slide. This is slide 118.
13 And, of course, there's the non-light 14 water reactor PRA standard, which was just issued last 15 year. This standard describes a full-scope PRA, which 16 includes consequence analysis, and there's a 17 mechanistic source term analysis element, or MS, which 18 provides useful information on what to do to develop 19 mechanistic source terms.
20 MS describes the objectives and the 21 characteristics and attributes of the mechanistic 22 source term analysis for the PRA element. It provides 23 information on what to do to develop mechanistic 24 source term, not specific assumptions, methods, 25 models, or computer codes. And it describes such NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
173 1 considerations such as timing, location, amount 2 released, and the radionuclide transport barriers and 3 transport mechanisms and associated uncertainties.
4 If an applicant is using the PRA standard, 5 I think there's good information in there that they 6 could reuse for their safety analysis beyond their 7 PRA.
8 Are there any questions on this aspect of 9 the PRA standard?
10 (No response.)
11 If not, I would like to hand the 12 presentation off to Bill Reckley to talk about the 13 Licensing Modernization Project.
14 MEMBER REMPE: So, I didn't realize you 15 weren't going to be the next presenter, Michelle. So, 16 I do have a question.
17 MS. HART: Yes. Sure.
18 MEMBER REMPE: Are you planning, is the 19 staff planning to endorse the INL or Sandia reports, 20 or something?
21 MS. HART: So, we haven't decided if we're 22 going to endorse them or not. Right now, we're 23 putting them out there as information that is useful 24 and could be used by a designer to help develop their 25 source terms.
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174 1 So, they weren't written in the sense that 2 they would be like a Regulatory Guide on acceptable 3 analyses to do a siting analysis, or anything like 4 that. They weren't written in that manner.
5 MEMBER REMPE: It just seems like you're 6 putting it out there, but it's not endorsed, or 7 there's no NUREG to say these are different approaches 8 that are acceptable to the staff. But if somebody 9 goes and follows that approach, and says, "Well, it 10 was on the website as something good to do, we 11 thought," and then, it's "No, no, it doesn't work."
12 So, I guess I'm surprised that they're not asking for 13 some sort of additional guidance. But, again, if 14 they're happy, I guess they're happy.
15 MS. HART: Well, I think the thing with 16 these, both of those reports, is that, much like with 17 the PRA standard, there's a lot to be left to the 18 implementation. It doesn't give specific models.
19 It's doesn't say, you know, "This is the way to do 20 things." It says, "These are the considerations that 21 you should use when you're developing a methodology."
22 And so, I think it's like a pre-step to 23 you determining how you're going to do that. And 24 then, you could come talk to us and say, "This is what 25 we're planning." And we're saying, okay, yes, if you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
175 1 use this information from these guides, we understand 2 where you're coming from.
3 MEMBER BROWN: How does that help an 4 applicant? I mean, to me, it would be just confusing.
5 I wouldn't know what would be acceptable and what 6 would not be acceptable.
7 This is Charlie Brown.
8 MS. HART: Right. I think --
9 MEMBER BROWN: You've gone through a whole 10 mess of possibilities, and, I mean, I'm confused. I 11 apologize.
12 MS. HART: So, they all have -- I'm sorry 13 -- they all have similar features. They all talk 14 about considering the same things. They all talk 15 about, you know, you need to quantify these things 16 -- "these things" being, you know, whatever they are.
17 I think there are some issues, but there 18 are so many different technologies; there are so many 19 different approaches that they can take to licensing 20 -- I mean, like they can determine to do a full 21 mechanistic source term or they can determine to do 22 something in between. Or they can determine to do a 23 maximum hypothetical accident, which has a non-24 physical, obviously, bounding analysis, where they 25 wouldn't have to do as much, I guess, justification of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
176 1 the assumptions that they had done, or detailed 2 analysis to supply the justification, is probably a 3 better way to say it.
4 So, I think we're trying to remain 5 flexible for these different approaches for these 6 different designs and different goals that these 7 applicants would have.
8 MEMBER BROWN: Well, one of them is 9 qualitative. Some of the other ones are more 10 quantitative.
11 MS. HART: Uh-hum.
12 MEMBER BROWN: It just seems --
13 MEMBER HALNON: So, this is Greg. Let me 14 try to frame it just a little bit.
15 These small reactors are technologies that 16 are usually -- I'll just say they're probably 17 commercial reactors or commercial facilities that will 18 be out there trying to make a profit. And in that, 19 there's going to be a tremendous economic pressure to 20 have extremely certain design parameters.
21 And when you get into these situations --
22 and I'm not sure if there's an answer to it or not --
23 where you get this sort of "bring me a rock, and we'll 24 let you know if we think it's okay," without that 25 bright line to design to, there's inefficiencies and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
177 1 there's potential economic issues.
2 So, there's going to be this drive to push 3 back. And I would hope that at some point we would 4 get, after enough technologies have come in, that we 5 could see the similarities in all of these and come up 6 with a process that gives more certainty to the 7 designers. That's the pressure, and there is 8 frustration amongst the designers because we don't 9 have, you know, how do you slice and dice the 10 licensing basis events, you know, get rid of the 10-11 to-the-minus-12 events and talk to 10-to-the-minus-5 12 events, and that sort of thing?
13 So, that's the frustration I think that at 14 least I'm experiencing through this. It came out in 15 droves when we were doing the 50.160 P rule. And I 16 think you're going through the same type of 17 discussions with NuScale.
18 MS. HART: Right. I think, you know, this 19 transition period we're in -- or I don't even know 20 that that's the right term -- but, yes, I mean, it's 21 hard to be efficient right out of the gate. We're 22 trying to be as flexible as possible and allow for 23 different concepts to come in. So, that does leave 24 some room for future refinements. And if we do see, 25 as we do see commonalities and obvious -- well, I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
178 1 won't even say, "obvious." As we learn lessons as we 2 go through, we may find in the future that there is 3 some guidance that we can give that could help refine 4 things and bring things into a more efficient space.
5 MEMBER HALNON: Well, I think it's 6 essential that we look for them and we have a project 7 in play that every time we do find one, that we can 8 refine the guidance, so that it gets more certain, we 9 should do that. We shouldn't wait five years, and 10 say, okay, now what did we learn? Each time we find 11 something that's going to be advantageous to the 12 certainty of a designer, we should allow that to be 13 used.
14 And I think the nature of Topical Reports, 15 and being able to use them across the board, is good, 16 except I keep on hearing, you know, this is very 17 specific to NuScale, or very specific to Kairos, or 18 it's very specific, and it causes a problem in the 19 transposition of these types of methodologies to other 20 reactors.
21 MR. SEGALA: This is John Segala with the 22 staff in NRR.
23 I just wanted to add on the question that 24 Joy had on the endorsement, I did want to add that, 25 for the non-light water reactor PRA standard, we're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
179 1 getting very close to putting out a trial use Reg.
2 Guide endorsing that standard. So, yes, we're not 3 endorsing the other two industry guidances, industry 4 documents -- I'm sorry -- the National Lab documents 5 that Michelle just presented on. Those are really 6 just being provided for information to help the 7 developers.
8 MEMBER REMPE: I get it, but it would be 9 nice if -- again, it's their bailiwick. If they're 10 not asking for it, I guess all you can do is say that 11 and be very careful to say, "These are just out here 12 for information. Use it. It doesn't mean that we're 13 going to approve it." But, anyway, that's what I'm 14 hearing.
15 MEMBER BROWN: Well, in my mind, this is 16 similar -- one of the concerns I've had with Part 53, 17 it's more generalized requirements as opposed to 18 general design criteria. And if somebody submits 19 that, then, all of a sudden, the NRC decides, well, 20 that's not quite good enough.
21 And the source term issue, it seems to me, 22 is going in the same direction: "Tell us what you'd 23 like to do and we'll figure out whether it's okay or 24 not." Flexibility is great, but that's really tough 25 to design hardware, plants, anything to get that work NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
180 1 done. It's almost like nobody wants to have any 2 specific requirements at all. I'm exaggerating 3 slightly, but that seems to be, if you're listening to 4 these slides, it seems to be where we're going. I'm 5 glad I'm not a designer.
6 So, I'll pass. I just think this is a 7 concern. I've spoken my piece.
8 CHAIRMAN PETTI: So, to me, my bigger 9 concern is the diffused nature of what's called, 10 quote, "requirements." They're there, but you've got 11 to go read six, seven, or eight documents -- and it's 12 only the source term, the most important thing we care 13 about in terms of protecting the public.
14 And that's why the idea of consolidating 15 it, you know, a website, whatever, so that someone 16 doesn't overlook something, (a) I think is good, but 17 (b) it would provide confidence, right, that, yes, the 18 NRC is doing its job and doing it well, and here's 19 what it looks like.
20 I would think NEI would be interested in 21 trying to put out something. Yes, it may be 22 functional, it may be higher level, because it goes 23 across technologies, but, as it is now, it is somewhat 24 diffuse. And this has to do with the NPUFs versus 25 power reactors. It has to do with the history of how, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
181 1 what's called the "source term tree" has grown, but it 2 seems like it's worth stepping back and asking these 3 sorts of questions at this juncture, given what the 4 future may hold in terms of advanced reactor 5 applications coming in.
6 Keep going, Michelle.
7 MS. HART: I'm sorry. Okay.
8 So, I will be handing off the presentation 9 to Bill Reckley now.
10 MR. RECKLEY: Okay. Thanks, Michelle.
11 So, thank you, everyone.
12 And I only have a few slides, but I'll go 13 back and visit a few of the comments, and not 14 disagreeing with anything anybody has said.
15 Part of the challenge is, as you provide 16 flexibility and try to make things technology-17 inclusive, addressing a variety of reactor 18 technologies, it gets harder to be specific and 19 produce anything like a TID document or even 20 NUREG-1465 and say, "Use this source term and it's 21 going to be used as a confirmation of a specific 22 barrier like the primary containment for a BWR or 23 PWR."
24 But, that said, just to kind of summarize 25 licensing modernization -- and we've talked to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
182 1 Committee and this Subcommittee a few times on this 2 -- it provides a risk-informed approach to the 3 selection of licensing basis events and categorizes 4 them into the categories we've talked about 5 -- anticipated operational occurrences, design basis 6 events, beyond design basis events, and then, another 7 category of design basis accidents that are used kind 8 of in a traditional deterministic method.
9 Those event categories are assessed and 10 compared to acceptance target figures to test specific 11 barriers and to assess the margin to the frequency 12 consequence target and design objectives, such as a 13 lower-dose value to justify an Emergency Planning 14 Zone, as Michelle talked about under the 50.160 15 rulemaking.
16 And this also includes an assessment 17 against cumulative risks, using the QHOs and some 18 other measures introduced by the NEI 18-04.
19 And then, just a summary that there are 20 key roles in the LMP, as we've talked about before, 21 for probabilistic risk assessment and, also, 22 mechanistic source term. So, I don't know.
23 If we can go to 120?
24 This is a figure similar to what Michelle 25 showed from the lab reports, and that's not a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
183 1 coincidence. We were all starting from the same 2 fundamental source document, which is a DOE order on 3 assessing radiological releases from any facility.
4 And so, it's simply modeled as an inventory, and then, 5 various barriers, and then, the integrity or the 6 performance of those barriers for particular scenarios 7 as it relates to particular radionuclides, and as it 8 relates to time, as the transients are being modeled.
9 So, Michelle had the same equation.
10 Obviously, we don't have it down yet to being able to 11 model things using an equation like this. But you 12 can, if you go back to Hossein's presentation and 13 Jason's presentation, you can see that these things 14 are what's being modeled in things like MELCOR, the 15 transient response of particular barriers and the 16 transport of radionuclides against those barriers.
17 So, one way to look at this, for example, 18 when Michelle talked about the Sandia report, in 19 looking at a first systematic way to look at it, it 20 would be, for a particular design, where am I going to 21 put the emphasis on retaining fission products? So, 22 under the current light water reactor model, a lot of 23 emphasis is put on barrier four, the containment 24 building. Under HTGRs, there's a much larger focus on 25 barrier one, the fuel, and maybe barrier two, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
184 1 matrix around the particles. In other cases, or for 2 other transient scenarios, since mechanistic source 3 term is done on a scenario basis, it might be the 4 pressure boundary or some other consideration.
5 And I know it's complicated. And I have 6 said, when it comes to helping the developers, if they 7 don't know it's complicated, they're getting into the 8 wrong business. They have to understand that, when 9 they're given these design choices and flexibility, 10 then they have responsibilities.
11 So, when you go to how they implement this 12 maze, if you will, let's say they are going to 13 emphasize barrier one in their design. That's where 14 they want to put the emphasis.
15 Then, Arlon, if you could do me a favor 16 and go back up to 115? Okay.
17 This is another complicated figure, but 18 what does it mean? It means, where are they going to 19 have to do their research? Where are they going to 20 have to put their emphasis? What does the R&D have to 21 go? Where are they going to try to minimize the 22 uncertainties in any particular barrier? If they're 23 putting the emphasis on the TRISO particle, there's 24 got to be a lot of work by DOE and by the developers 25 to show that they've addressed that qualification of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
185 1 the TRISO fuel. On the other hand, should they choose 2 to say, "I'm going to put my emphasis on another 3 barrier," then, they're going to have to do the R&D on 4 that other barrier.
5 So, again, it is complicated and it's 6 iterative, how a designer is going to have to walk 7 through this. But, again, given that we're trying to 8 address a lot of different technologies, and we're 9 trying to give flexibility to the designers as to 10 where they put the emphasis in their design and in 11 their analysis, it's just the nature of the beast that 12 it won't be as clear as giving them a spreadsheet with 13 radionuclides and saying, "Here, put this into this 14 volume and model your containment."
15 And I oversimplified. I don't mean to --
16 even the current process is not as simple as that, 17 but, anyway --
18 So, Joy, I see you've got your hand up.
19 MEMBER REMPE: Yes. You're right, and 20 what you just said would be good to have in a document 21 somewhere.
22 CHAIRMAN PETTI: Amen, amen, amen.
23 MEMBER REMPE: The other thing I have to 24 say is --
25 CHAIRMAN PETTI: I completely agree.
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186 1 Bill, it was good.
2 MEMBER REMPE: Let me finish, Dave. Let 3 me finish, Dave. Okay?
4 CHAIRMAN PETTI: Okay.
5 MEMBER REMPE: Old farts like me can 6 recognize this equation really comes from NUREG-1150.
7 This is not something new. But perhaps the new design 8 developers are not old farts like myself. Okay?
9 (Laughter.)
10 So, yes, and what you're saying is very 11 good, Bill, and as Dave is saying, "Amen," and singing 12 a chorus here. It ought to be written somewhere. Is 13 it written somewhere?
14 MR. RECKLEY: I think the reports that we 15 commissioned talk about it in this sense, when you 16 look at them. And, for example, the Idaho report, 17 then, goes on to further say things like Michelle was 18 describing. You have an option in there at some point 19 to say, "I'm going to address the uncertainty by being 20 conservative. There's things I don't know in terms of 21 the behavior of the radionuclide, or the behavior of 22 a barrier. I'll just be conservative," and that's an 23 option that's available to the designer as well to get 24 through the licensing process.
25 So, whether it's as clear as it should be, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
187 1 I'm not sure. As John was saying, one thing we can do 2 -- or even some of the members were saying -- bring 3 this up at a public meeting and ask the developers, 4 who are the ones we're trying to do this for, is it 5 clear; is this understandable? Between the website, 6 the reports, the history, is it clear? I don't know.
7 MEMBER REMPE: Okay. I think you've got 8 our point. But Dave may want to expand --
9 CHAIRMAN PETTI: No, no, I agree. That's 10 why I keep thinking of this, sort of an executive 11 summary that pulls all this together. I'm thinking 12 three-, four-, five-page manual, heavily annotated.
13 So, it's kind of a roadmap, but the important things 14 get put there, so that it helps them put all these 15 different documents in context.
16 Because, you know, it is diffused right 17 now, and I'm just trying to figure out, is there a way 18 to better focus it, just to help them? I agree with 19 you that you still need the flexibility, and it's 20 never going to be as cut and dry as "Do X, and then, 21 do Y, and then, do Z." But if we can make sure that 22 they know what the considerations are in each step, 23 and what's important, that's, I think, the best you 24 can do at this point.
25 MR. RECKLEY: Okay. Well, I guess -- and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
188 1 John will pick this up maybe at the end with the 2 website -- that's a step; that's our effort to take a 3 step in that direction. And you guys haven't had a 4 chance to look at it yet, and neither have the 5 external stakeholders. Maybe the next step is to see 6 if that provided additional clarity where we're 7 seeking it.
8 So, Arlon, if you can go back up to 120?
9 MEMBER REMPE: While you're going up 10 there --
11 MR. RECKLEY: Yes?
12 MEMBER REMPE: -- again, I'm thinking 13 about websites and how they can easily be changed.
14 And it's nice to have the website, but you always have 15 to say, "Last accessed," and what date you did, 16 because --
17 MR. RECKLEY: Yes.
18 MEMBER REMPE: -- it can easily be 19 changed. And so, again, I think something that's a 20 little more concrete might be a good idea.
21 MR. RECKLEY: Okay.
22 MR. SEGALA: Yes, this is John Segala, 23 Bill.
24 I just wanted to add, I mean, now that we 25 made the website, the web page on source term, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
189 1 publicly available just earlier this week, you know, 2 we were planning to share that at an upcoming 3 stakeholder meeting, just to make everybody aware that 4 we have this new page. So, just throwing that out 5 there.
6 MR. RECKLEY: Okay. And, Arlon, if we can 7 go to 121?
8 It's basically a very similar 9 representation, just coming at it from a different 10 direction. And as I just mentioned, and the bullet 11 emphasizes here, there is flexibility provided on how 12 they want to develop the safety case, both to reflect 13 the design and within the analysis.
14 And that involves combinations of active, 15 engineered safety features, passive safety features, 16 and increasingly, reliance on inherent properties of 17 materials in the reactor cores. And that's 18 consistent, as it says here in the second bullet, with 19 the Advanced Reactor Policy Statement encouraging 20 passive and inherent features.
21 But sometimes, those also bring in 22 uncertainties and things you need to address in both 23 the design and in the analysis.
24 So, that's just a quick summary, again, of 25 licensing modernization; the key role that modeling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
190 1 the mechanistic source term has in that.
2 So, with that, if there's no other 3 questions or suggestions, we can go on to fuel 4 qualifications and Tim.
5 MEMBER KIRCHNER: Dave, may I make an 6 observation? This is Walt Kirchner.
7 CHAIRMAN PETTI: Sure.
8 MEMBER KIRCHNER: You know, Dave, and 9 going back to this morning, we have this DG -- let me 10 make sure I get the number right -- DG-1199. Is that 11 an appropriate place to at least lay out this generic 12 approach to source term analysis, identification and 13 analysis?
14 CHAIRMAN PETTI: I think that's RWR-based.
15 MEMBER KIRCHNER: I know it is, but, 16 conceptually, it seems to me, at least what I do is, 17 each time we see a new document for non-LWRs, my first 18 proof of concept or test of the document is, would it 19 work for an advanced LWR? And I think they will, most 20 of them.
21 And the advanced LWR, of course, is where 22 we've got the most experience. We've got very mature 23 PRAs. We've got equipment reliability to feed into 24 the PRAs. We've got a good understanding of the 25 technology and vulnerabilities, et cetera, et cetera.
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191 1 So, you know, in my mind, when we look at 2 these new approaches, I'm testing them on an advanced 3 LWR, and I think they will work. So, conceptually, 4 the charts we were just looking at on mechanistic 5 source term, I mean, those could be used for an 6 advanced LWR as well.
7 I thought what Michelle was covering, that 8 synopsis of the methodology from the INL report was 9 useful because you started off trying to bound the 10 problem, and then, you worked further down. If you've 11 got the data, if you've got the experimental data, if 12 you've got the experience with the technology, you can 13 probably go down further. As Bill was saying, this is 14 complicated, especially for newer technologies where 15 you don't have the kind of databases, and such, and 16 experience.
17 It just seems to me that this DG-1199 18 might be a good place to put the conceptual approach 19 in, in a technology-inclusive manner, and then, go on 20 to --
21 CHAIRMAN PETTI: My view is, if there was 22 a place that it needed to be put, it would be in an 23 appendix to the NUREG on fuel qualification that Tim's 24 going to talk about. Sorry, Tim. To me, that was 25 where it made most sense. But I would leave that up NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
192 1 to the staff to decide where to put it. But there's 2 a lot of --
3 MEMBER KIRCHNER: But that's just fuel.
4 To do the source term --
5 CHAIRMAN PETTI: No, the source term is --
6 MEMBER KIRCHNER: -- you've got to have 7 the barriers.
8 CHAIRMAN PETTI: It's talked about in 9 there. You'll hear it. You'll hear it from him.
10 That stuff is in there. But, again, I'd leave those 11 details sort of to the staff, but --
12 MEMBER KIRCHNER: That was just a comment.
13 No need -- no answer required.
14 MR. DRZEWIECKI: Okay. So, this is Tim 15 Drzewiecki from the staff.
16 Sorry, can you guys hear me?
17 CHAIRMAN PETTI: Yes, go ahead.
18 MR. DRZEWIECKI: Okay. Wonderful. Okay.
19 Yes. So, yes.
20 So, I've got a few slides to just kind of 21 give an idea of what's in this document. This is the 22 fuel qualification guidance that we came out with and 23 we talked about at the end of last year.
24 And what it kind of lays out is a top-down 25 method to kind of highlight a list of goals or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
193 1 criteria the staff could evaluate to say that this 2 fuel is qualified and it does have certain aspects 3 that I'm going to touch on source term. So, I'll kind 4 of walk through how it relates to the source term in 5 the next couple of slides.
6 So, may I have the next slide, please?
7 So, how it starts out is we're trying to 8 make a finding that the fuel is qualified for use.
9 And there are two kind of conceptual things that we 10 want to make findings on for that. One, that you have 11 the manufacturing specification to control the key 12 fabrication parameters, and the other puts that the 13 safety criteria can be specified. Obviously, the 14 safety criteria are not well-defined. So, we break 15 that down more on the next slide.
16 So, the next slide, please.
17 And in here, specific criteria are coming 18 from the NRC regulations. The box on the left, that's 19 talking about margin on the safety limits under normal 20 operation and the AOOs. And the one in the middle, 21 which is the focus of this presentation, is showing 22 that you can maintain the margin of a radionuclide 23 released when it's under accident conditions.
24 Now the point is, for this report, we were 25 focused on the fuels role, but it does talk about NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
194 1 other barriers, too. And this goes into the role that 2 fuel plays in the protection against the release of 3 radionuclides.
4 And the box on the right, that just goes 5 into, basically, showing that you can maintain a safe 6 shutdown condition.
7 Next slide, please.
8 Okay. And so, that box that was in the 9 middle, it's broken down into four other criteria. I 10 want to say that the boxes that are gray are gray 11 because they're considered these base goals, in the 12 sense that we would expect to be able to have enough 13 evidence to make a finding on that item and not have 14 to break it down any more.
15 So, in terms of the box on the left, we 16 want to make that the fuel performance envelope is 17 defined. In other words, for this goal, it's knowing 18 what kind of accidents that the fuel is going to be 19 credited to perform under. So, whether it's under-20 cooling events, overpower events, things like that.
21 The box second to the left is, basically, 22 specifying what's the radionuclide retention 23 requirements of the fuel under accident conditions.
24 This is going into the role of fuel as it plays in the 25 safety case, in the sense that, if you're not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
195 1 crediting your fuel, if you have other barriers, 2 that's going to have an impact on the kind of data 3 that you would need to support that.
4 These other two boxes on the right, these, 5 I think, line up pretty well with the source term, a 6 slide that was shown on -- both Bill and Michelle had 7 shown it. So, these are associated with, one, the 8 barriers to the release of radionuclides, showing that 9 you're conservative on criteria there. So, this 10 should be things like propellant/clad/mechanical 11 interaction, your limits, or it would be other things, 12 too, like whether it's, say, like melting, and things 13 like that.
14 And then, the box on the right, that's 15 talking about the release of radionuclides or the 16 migration of radionuclides within the fuel matrix.
17 So, I was going to break down the box, 18 that whole G2.2.2 on the next slide. Yes.
19 Essentially, this has two boxes. One is 20 that you can show the criteria are conservative, which 21 we think is something that can be shown, as long as 22 you have data. This assumes that you're comparing it 23 against experimental data.
24 Then, the box on the right is just making 25 a statement that experimental data is appropriate.
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196 1 And that box is blue because there's a whole separate 2 framework that's going to break that down. And that 3 goes into things like data independence. Is that data 4 applicable to the scenario, things like that?
5 And so, what I have on this next slide 6 -- may I have the next slide, please? -- so, this is 7 not really meant to be read, but it kind of shows you 8 what this framework is. It takes a high-level goal 9 and it breaks it down into, effectively, like a 10 checklist, a bunch of things that we want to make 11 findings on. And if we can make findings on all of 12 those gray boxes, then we can say that the fuel is 13 qualified.
14 And there is some overlap here. And what 15 I mean by that is the main framework is the one on the 16 left, and that relies on things like having evaluation 17 models in a couple of sections and experimental data 18 is used to support multiple goals as well.
19 So, that's all that I had for NUREG-2246, 20 unless there's any questions.
21 (No response.)
22 Arlon, can you hear me? Okay.
23 MR. COSTA: Yes, we can hear you.
24 MS. HART: I guess if there are no 25 questions, we can go to the next slide.
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197 1 So, I wanted to talk a little bit about 2 this web page that we keep referring to, the non-light 3 water reactor accident source term web page 4 information. At the top, there's the link to it.
5 It's on the public website and it's underneath the 6 advanced Reactor-related documents.
7 And it's a one-stop shop for existing 8 information, and it's discussion of accident source 9 terms. There's a link list of documents relevant to 10 development of non-light water reactor accident source 11 terms for licensing. It includes some of the light 12 water reactor information. We're not trying to be 13 comprehensive. Right now, we are putting information 14 out there and we will keep it up-to-date and keep 15 adding to it, as we continue to go through this 16 process.
17 Arlon, if you could open the website, so 18 they can take a look at it?
19 So, we have a little bit of a preamble 20 that talks about what a source term is and a little 21 picture of the barriers and kind of the barriers 22 assessment.
23 And next, we go through the history and 24 evolution of the light water reactors source term.
25 So, you have some of that history.
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198 1 And then, it kind of follows along the 2 lines of what we've been talking about today. We do 3 have the links in the text here to TID-14844 and Part 4 100, Reg. Guide 1.183.
5 We talk a little bit about the analytical 6 tools and past studies. So, we have references to 7 RADTRAD and MELCOR.
8 And then, we talk about information on 9 SMRs and non-light water reactors. So, we do have a 10 discussion to our mechanistic source term and how it 11 interacts with the regulatory analysis and the siting 12 analysis that we talked about in SECY paper 16-0012.
13 And then, we talk about the SECY-93-092, 14 which was kind of the base for the information on what 15 a mechanistic source term is and the considerations 16 for what the staff would look at for a good 17 mechanistic source term.
18 And then, further down on the page, we 19 have a section where we would talk about accident 20 consequence-related regulation activities. We don't 21 have links to those right now at the moment, but it 22 says that there are things that we're going through 23 now.
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199 1 term with reference to the LMP, is the first thing, 2 the Regulatory Guide that endorses the LMP.
3 And then, we go through document links.
4 And so, you can see that we have a link to TID-14844, 5 NUREG-1465, Reg. Guide 1.183. And then, Hossein's 6 discussion about the history of source terms.
7 We have a link to our vision and strategy 8 code development plans for severe accident analyses.
9 We have some links to those demonstration 10 projects that Hossein and Jason discussed earlier 11 today. So, we do have the slides and a video 12 recording of those presentations from the workshops 13 available on this site.
14 A little bit further down, another link to 15 SECY-94-302, which is about source term-related, you 16 know, the information on source terms for evolutionary 17 and passive light water reactors. So, it's not non-18 light water reactors, but it's that kind of 19 information.
20 And we have the wrong SRP here. It should 21 be 15.0.3 for new reactors. We can fix that.
22 We do have a link to the approved NuScale 23 Topical Report for Accident Source Term Methodologies.
24 So, if somebody wanted to see an example of an 25 approved methodology for light water reactors SMRs, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
200 1 and we do have the SECY, our staff approach to 2 evaluating the source terms for NuScale.
3 And then, here, we have some SECYs and 4 some NUREGs and other information on the source term 5 approach for non-light water reactors, including 6 SECY-93-092 and it's SMR; some pre-application Safety 7 Evaluation Reports for several of the designs, and 8 then, an assessment of a white paper from NGNP on fuel 9 qualification and mechanistic source terms. So, 10 that's our staff assessment.
11 In the next section, we have some guidance 12 on developing advanced reactor source term. And so, 13 for that, we have the endorsement Reg. Guide for 14 NEI 18-04.
15 We do have the Reg. Guide for the 16 performance-based emergency preparedness for small 17 modular reactors and other new technologies.
18 And then, we have the lab reports that I 19 described today and a link to the ANS standard, which, 20 of course, we are working on an endorsement Reg. Guide 21 for that as well.
22 And then, the next section has several 23 reports from labs and from other places about source 24 term information on designs that we've seen in the 25 past.
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201 1 And then, the next section is 2 presentations on non-light water reactor source term, 3 and I expect that we will probably be putting this 4 presentation and transcript, a link to that here as 5 well.
6 Like I said, this is what we put so far.
7 And we do intend on keeping that up-to-date. And I do 8 like the idea that Joy was talking about, that maybe 9 we need to have a "last updated" note on there.
10 CHAIRMAN PETTI: Yes, the other thing you 11 could do is, if something's new, sometimes like 12 websites put a little tag, "New" next to it.
13 MS. HART: Uh-hum.
14 CHAIRMAN PETTI: So, it draws the eye of 15 people visiting it.
16 MS. HART: Yes.
17 CHAIRMAN PETTI: I'm just wondering, 18 nothing is there about NUREG-1537 for the NPUFs, and 19 that might be useful, to consider bringing in some of 20 those documents, too.
21 MS. HART: Right. And I think that's a 22 good piece of information. I think there's also been 23 some discussion about maybe some information that they 24 had used when they developed the source terms for like 25 SHINE, or whatever. We can look into what information NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
202 1 we can include on that.
2 CHAIRMAN PETTI: Right, right. I think 3 that would be good.
4 And, I mean, I like that introduction 5 stuff. That's what I was kind of thinking of as 6 needed. So, I think it's a good flow of all that.
7 MS. HART: Yes. And it's a very high 8 level right at the moment. As we go through learning 9 things again, you know, and as we get some comments 10 from folks as they look at it, not that we're 11 requesting comments officially, or anything like that, 12 but if there's missing pieces of information or 13 information that doesn't make sense, certainly, we'll 14 take that comment into consideration.
15 MEMBER REMPE: The intro is good, but some 16 of the things that Bill mentioned about where you 17 emphasize and the barrier diagram might be good to 18 consider --
19 MS. HART: Uh-hum.
20 MEMBER REMPE: -- in the actual formation, 21 the introduction. But, again, that's just one 22 member's comment.
23 MS. HART: Sure.
24 MEMBER REMPE: And we're getting into 25 designing your web page, which I don't think ACRS NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
203 1 members should be doing. But it's just something to 2 think about.
3 MS. HART: Right. No. No, I understand.
4 You know, there could be more pictures and more 5 discussion. Or they could be more callouts to other 6 documents, as necessary.
7 So, as we go through this, and if we hear 8 from industry or anybody about what they think is 9 useful, we can take that into consideration. I can't 10 make any promises necessarily, because we're not 11 asking for guidance on how to do our website 12 necessarily, but we thought that this would be a good 13 way to gather this information. And it may be a 14 little bit more flexible than trying to write like a 15 series of white papers or SECY papers, or something, 16 which takes a lot of time and effort to try to get all 17 the language correct.
18 MEMBER REMPE: And I do want to say, I 19 were a developer, having all the documents in an 20 accessible place would help big time, and sharing your 21 knowledge this way is good. It's just it's very 22 fluid, I guess. I'm still thinking that some sort of 23 paper document might be good to be thinking about.
24 MS. HART: Yes, we can also think about 25 this as we go through things. I will say, you know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
204 1 a certain amount of what is on this web page is 2 constrained by the way that the NRC website is 3 generally configured. So, there may be some of these 4 good ideas that you have that maybe aren't 5 implementable or generally haven't been implemented on 6 our website. So, I'll just add that as a thought to 7 throw out there.
8 MEMBER REMPE: Good point.
9 MS. HART: Are there any other questions?
10 DR. BLEY: Michelle, yes, this is Dennis 11 Bley.
12 After you introduced it, I went over and 13 found it and have been poking around in it. And I 14 think it's a great start. And it's the kind of thing 15 I think the Committee has been looking for for a 16 while, and I think it will be very helpful to people.
17 Thanks.
18 MS. HART: Right. And somebody did tell 19 me that there is a timestamp on the page near the 20 bottom that says the last time it was updated. So, 21 it's not in big print or anything like that, but it 22 does have it.
23 MEMBER REMPE: It is on there. But I 24 guess I was thinking, if a person came in and said, 25 "Well, I used document X that was available on January NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
205 1 15th, 2021," or 2022, and in the ensuing two years 2 before they came in, that document has been removed 3 because the NRC has decided that wasn't a good idea to 4 post it, they're kind of stuck. If you say, I used 5 Reg. Guide 1.-whatever that was Rev. 3, the fact 6 you've updated it, well, it was there at that time and 7 there's more certainty with the document.
8 MS. HART: Right. I think, right now, we 9 only have things that already existed. We haven't 10 added new, different information, other than this 11 introduction/preamble stuff. So, if you're using a 12 link that's on the page, you should be able to find 13 that again later, if you've taken down the 14 information. Most of them are ADAMS documents; not 15 all of them are, though.
16 And, of course, there are some pluses and 17 minuses to that as well. I'm thinking of like 18 specifically Reg. Guide 1.183. Right now, we have a 19 link to Rev. 0. What happens when we have Rev. 1? We 20 have to make sure that that's up-to-date.
21 MEMBER REMPE: And again, these are just 22 comments, and you've emphasized ACRS members shouldn't 23 be designing a website. But it's just, you know, we 24 can't help ourselves sometimes.
25 MS. HART: No, I get it.
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206 1 Are there any other questions or concerns?
2 (No response.)
3 All right. If not --
4 DR. BLEY: Well, yes. Now that you 5 mention that there are constraints on NRC websites, 6 I'm reminded of the briefings we've had on NRR's 7 Venture Studios, and maybe they can be of help to you 8 guys. And maybe you've already used them to make this 9 more flexible and easier to use, and flexible in areas 10 where you're looking for more flexibility.
11 MS. HART: All right. Thank you.
12 So, if there are no further questions, I 13 guess we can move on to the next topic, and that will 14 go to Bill Reckley.
15 MR. RECKLEY: Okay. Thanks, Michelle.
16 And I think the only reason I'm doing this 17 is because I was gone in January. So, I stepped 18 forward or everyone stepped backwards, or however that 19 works.
20 So, just kind of summarizing things -- and 21 I'll ask Michelle or John to jump in whenever it's 22 appropriate -- but if we want to go to 131, Arlon?
23 Just kind of, as a summary and a general 24 approach, using that same kind of complicated figure 25 that Michelle used, but it does emphasize the level of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
207 1 effort that needs to go into this.
2 And we understand the general observation 3 that reactor developers, and especially those that may 4 be somewhat newer to this kind of interaction with the 5 NRC, it's complicated and maybe it could be made 6 simpler. On the other hand, this is really what it 7 comes down to, right? I mean, this is, as I think 8 Mark said in the very beginning, this is why the NRC 9 is here, and that's to address the potential hazards 10 of nuclear reactors and radioactive materials, and 11 ensure that they aren't released to the environment.
12 And so, as you look at how that evolved 13 for light water reactors since the 1950s or the TID in 14 1962 to present day, that reflected the way that it 15 was done, the importance of the containment, all the 16 way back to the genesis of that, as part of the siting 17 decisions of the Atomic Energy Commission. And it's 18 worked over that period of time.
19 And so, as the first bullet says, if a 20 developer wants to use that kind of an approach, if 21 they're a non-light water reactor, they're going to 22 have to show a conservative source term, and it may 23 not be the same as NUREG-1465 because of the 24 technology differences. But if they want to put an 25 emphasis on a barrier like a containment and use that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
208 1 same approach, that remains available to them.
2 The other thing that the second bullet 3 starts to go into is they can, also, decide to use a 4 more mechanistic source term and a more LMP-type 5 approach, where you are assessing the performance of 6 barriers and the behavior of radionuclides for 7 particular scenarios and assessing the system overall.
8 That's the functional containment concept. You're, 9 basically, looking at all of the barriers and their 10 performance for various scenarios, as the LMP would 11 allow you to do.
12 In either case, the third bullet, when 13 you're dealing with a variety of technologies and 14 designs, the actual implementation is going to be 15 specific to that technology and that design. The 16 source term for a molten salt reactor, and what 17 radionuclides stay in the salt and what gets released, 18 and what barrier they're going to rely on, is going to 19 be different than a gas reactor or a sodium fast 20 reactor. And that will go to my next slide to some 21 degree.
22 The last bullet is the NRC is not 23 currently planning to do the equivalent of a TID 24 assessment or a 1465, where we give source terms to 25 applicants and say, "If you use this, it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
209 1 acceptable."
2 As Jason and Hossein mentioned, the work 3 we do, we share, but that's somewhat different than 4 giving them something in a Regulatory Guide and, 5 basically, saying, "Use this as an input." And the 6 reason we're not planning to do that is for a couple 7 of reasons. One, the variety of technologies, and 8 two, a general trend where that kind of scientific 9 work, the burden of doing that is put over onto the 10 applicants.
11 And if you want to go to 132, Arlon?
12 Obviously, the Department of Energy and 13 the National Laboratories play a role in this. And 14 so, as you know, the National Labs are organized into 15 various campaigns for different technologies. All the 16 labs associated with taking the lead for a particular 17 technology are looking at the behavior of 18 radionuclides and barriers and materials, and how that 19 plays into source term.
20 I just threw up a couple of reports. The 21 HTGR, that dates back from the NGNP, Next Generation 22 Nuclear Plant, project from 10 years ago. Argonne and 23 Oak Ridge are doing work on molten salt. Argonne is 24 doing work to develop models for sodium fast reactors.
25 So, if we go to 133.
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210 1 The NRC's activities have been more 2 focused on the methodology and how to use that 3 scientific data, or how an applicant might use that 4 scientific data, that they do by their own experiments 5 or through reference to DOE activities, to support 6 their design and to support their safety case, again, 7 going back to the previous discussion of where they're 8 going to put their emphasis. And both of these 9 reports kind of echo the flexibility that designers 10 have and where they might want to put their emphasis.
11 Joy?
12 MEMBER REMPE: So, I'm hearing a little 13 bit different from you in the prior slide discussion, 14 which maybe you're right, but I also think that the 15 design developers need to understand that, too. I 16 still think I like the discussion about, if you rely 17 on that barrier, you'd better have more research.
18 That's technology-independent, and I think that to be 19 vetting guidance.
20 But you said, I believe -- and I'm 21 paraphrasing -- that, because we've got so many 22 technologies coming in and they're not using the 23 standard LWR stuff, the burden is on the new design 24 developer to come up with a source term. That's just 25 the way it's going to be. If you're going to pick a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
211 1 new technology, you've got to realize you've got to 2 have enough money to get the data and the approach for 3 your source term. And maybe that ought to be in some 4 guidance. Or is it communicated well to them, so they 5 understand that?
6 MR. RECKLEY: Well, yes.
7 MEMBER REMPE: Am I misquoting you, by the 8 way? Because I am kind of paraphrasing what I'm 9 hearing.
10 MR. RECKLEY: No, no. I think it's an 11 accurate paraphrase. The only thing -- Arlon, if you 12 can go back to 132? -- is, and I think you guys are 13 aware of this, all of these are complex relationships, 14 right?
15 The Department of Energy is not an 16 innocent bystander. They are a key player. And they 17 do this through their own work. They do it through 18 working with developers. They solicit input from the 19 developers on where to do the research through things 20 like Project GAIN. They work with individual 21 developers, and some of the Project GAIN grants have 22 related directly to developing source terms.
23 And we're not just standing by, either.
24 We're interacting with DOE on where to do some of the 25 research. And so, you do have all parties kind of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
212 1 trying to stay cognizant of what each other is doing 2 to make sure this works out. I don't want to make it 3 sound like we're just stepping back. We're talking to 4 the developers, as Michelle and others have said, in 5 this pre-application discussions. Those same 6 developers are talking to DOE. We're talking to DOE.
7 And so, the hope is that, although it is complicated, 8 that it is kind of an organized approach to this.
9 So, I'll leave it there, unless John or 10 Michelle want to weigh-in on the interactions we have 11 with the developers and the laboratories and DOE.
12 Okay?
13 MEMBER REMPE: It helps. Again, it's just 14 something I'm thinking about. I'm reacting to things 15 I see in the popular press.
16 MR. RECKLEY: Right.
17 MEMBER REMPE: And I'm thinking that some 18 points are worth emphasizing.
19 MR. RECKLEY: No, and I don't think any of 20 us really have much different thoughts. I mean, in 21 terms of the development of a technology, it's like a 22 3-dimensional chess game here -- working with the 23 developers, the DOE, the NRC, and other entities. As 24 Hossein was mentioning, you can bring in the 25 international elements. It's a complex set of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
213 1 activities to support any of these technologies.
2 MEMBER REMPE: Go ahead. Thank you.
3 MR. RECKLEY: Sure.
4 So, Arlon, if we want to go to 134?
5 You know, this is just kind of reinforcing 6 that complexity. This is maybe the most famous 7 simple, well, single figure representing a source 8 term, and it goes back to the NGNP and HTGR 9 discussions.
10 But, if you look at those DOE reports, you 11 will see similar representations for molten salt 12 reactors, similar representations for sodium-cooled 13 fast reactors. And you can do this for any of the 14 designs and the technologies. And it's just another 15 way to try to represent sort of what was going into 16 the barrier diagram, and just another way to try to 17 represent mechanistic source term; and, again, the 18 complexities of trying to, for individual scenarios, 19 say which of these is going to come into play; which 20 of these are going to contribute? In this particular 21 case, when are the contaminants going to come off of 22 the primary circuit and get released into the reactor 23 building? When might they be retained, and so forth?
24 So, I didn't really have much of a 25 message, other than just using this somewhat familiar NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
214 1 capsule diagram to kind of reinforce the number of 2 things going on and the complexities of developing a 3 source term.
4 So, go to 135.
5 This is just, again, summarizing what 6 Jason and Hossein were saying. Our activities in 7 regards to this, developing the models, which we 8 share, for at least the consideration of the 9 developers, whether they're using MELCOR or some other 10 system analysis code to support the design and 11 licensing of their plant.
12 So, 136.
13 Again, this is, to some degree, just 14 summarizing the previous discussions, as Michelle went 15 through. We have had both applications and pre-16 applications with NuScale, along with Kairos, and 17 then, pending discussions with Westinghouse, 18 TerraPower, X-energy. And we'll consider those in 19 terms of a step forward; what is useful to share.
20 As another developer comes in, we can talk 21 to them about at least the public versions of the 22 interactions with these other vendors, to say, you 23 know, this is one approach that was taken. And if 24 it's an approved Topical, then so much the better; 25 that this approach was found to be acceptable.
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215 1 Slide 137.
2 So, moving forward, we, the staff, will 3 continue to follow the scientific work being done by 4 the National Labs and the developers. We continue to 5 engage with those developers on the approaches they're 6 taking in both the design and for presenting a safety 7 case.
8 We'll continue, as Hossein and Jason 9 mentioned, to develop, refine what we've done to date, 10 and to develop models for the last two technologies we 11 plan to do.
12 And in terms of the MELCOR workshops, we 13 will engage stakeholders, including the meeting here 14 today, and consider additional guidance, if we get 15 feedback like the last bullet: the web page maybe was 16 useful. But if stakeholders are commenting that it 17 needs to either be revised, we'll take that into 18 account, or we need to do something more, then we'll 19 consider that, based on the feedback we get from our 20 engagement with stakeholders, individual developers, 21 and so forth.
22 So, I think that's the last slide.
23 Questions?
24 MEMBER HALNON: Bill, this is Greg.
25 Given that last bullet there, maybe an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
216 1 inclusion on the web page would be a link that emails 2 a certain person that could take those kind of 3 comments. I don't know if that's possible in the 4 public world, but it seems it would be an easy way to 5 get some feedback on it, rather than just wait for 6 public meetings or someone to get so frustrated that 7 they call you. It might be a good way to get some of 8 that feedback.
9 MR. RECKLEY: All right. We'll put 10 Arlon's name right on there.
11 MEMBER HALNON: That's what I was 12 thinking.
13 MR. COSTA: Appreciate that.
14 DR. BLEY: Hey, Bill, it's Dennis Bley.
15 I haven't looked up what's on the public 16 web about Part 53 lately, but have you thought about 17 doing anything similar to support Part 53?
18 MR. RECKLEY: We have quite a bit on Part 19 53. I think this one might flow a little better than 20 our current Part 53. So, we'll take a look at it.
21 But we do have a lot of information on Part 53 22 currently on the public website.
23 DR. BLEY: And I was just thinking about 24 the roadmap, and we've talked about that before.
25 MR. RECKLEY: Yes.
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217 1 DR. BLEY: But that might be a place to do 2 that.
3 MR. RECKLEY: Okay. Thank you.
4 Hey, John, I'm not sure if you had any 5 additional kind of --
6 MR. SEGALA: Yes, this is John Segala.
7 I mean, I appreciate all the discussions 8 and the feedback, and I appreciate the opportunity 9 that we had today to present on all the activities 10 that we've been working on, and the new web page, 11 everything.
12 I think I'll let you guys conclude and how 13 you want to move forward. But I guess I'm just kind 14 of hearing that some of the things that we're thinking 15 about as sharing the new web page at our upcoming 16 stakeholder meeting in March, you know, reaching out 17 to stakeholders at that meeting and see if there's a 18 need for something else in terms of source term.
19 It seems like, from what I've heard, that 20 there seems to be a general flavor that there's a lot 21 of good information out there. It's just, you know, 22 there's a lot of information scattered. So, it's hard 23 for a new developer to look at in more of a 24 consolidated manner.
25 Our attempt with the website was an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
218 1 attempt to try to pull it all together and make it 2 easy for people to find. And as Michelle and Bill 3 said, we're going to continue to refine that web page 4 and enhance it, and add to it as new information comes 5 out.
6 So, anyway, again, I appreciate the 7 discussion and even Joy saying that ACRS providing 8 feedback on our websites -- we're open; any sort of 9 feedback for things that we can do better to 10 communicate externally, you know, that's something 11 that we're interested in.
12 With that, I think that sort of concludes 13 the staff remarks.
14 MEMBER REMPE: Again, it was just 15 individual member comments. This is a work-in-16 progress. John, do you think it's appropriate, I 17 mean, knowing that this is a work-in-progress, do you 18 think an ACRS letter is really going to help the 19 staff? You've got our individual comments.
20 But I heard you say at the beginning of 21 this meeting, "We hear there's going to be a letter."
22 You guys were not asking for a letter. What's your 23 thought on this? I mean, we're doing individual, fly-24 by-night comments here. Do you want us to be more 25 measured and come up with a letter? Or do you think NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
219 1 you've got enough?
2 MR. SEGALA: Well, I think, from our 3 perspective, we feel like we have enough for 4 developers. We feel that the early-moving developers 5 that were engaging with or developing source term, we 6 haven't really heard any strong interest from the 7 developers through our interactions with them in pre-8 app, or even in the public meetings, that there's a 9 need for additional guidance.
10 And so, we kind of feel like we don't 11 really need a letter. We tried to listen to the 12 feedback that you all provided to us at earlier 13 meetings on EP and other things. And we've tried to 14 do all sorts of activities, like develop guidance and 15 develop the MELCOR-SCALE demonstration projects, and 16 make the videos and information available, and the 17 contractor reports.
18 So, I think what we decided sort of at the 19 end was, you know, that that's where we kind of ended 20 up with a web page. We thought, well, maybe we do 21 need to pull all this information together and make it 22 easy access. And so, that was sort of what we thought 23 was kind of the piece that was needed to kind of pull 24 it all together. And it is a work-in-progress, and 25 we're going to continue to engage with these NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
220 1 developers and applicants, as they submit applications 2 and do the reviews.
3 You know, we've been developing guidance, 4 as you can see from this whole presentation, we've 5 been developing guidance for LWRs since the early 6 '60s, and we continue to develop guidance for LWRs.
7 And I imagine that we will continue to do that for 8 non-light water reactors, as well as SMRs and other 9 new technologies that come forward.
10 So, I think we're trying to remain 11 flexible. We're trying to remain agile. We're trying 12 to optimize and enhance our guidance, as we move on 13 and as we learn new information.
14 So, I'm not sure if that answered your 15 question, but that's kind of our view.
16 MEMBER REMPE: It helps to understand.
17 Again, we do this now with the research folks; that 18 we're having more frequent meetings, which takes more 19 staff time to come and talk to us, and individual 20 members provide comments. And we hear from your 21 personnel that's more helpful to them than having 22 formal letters coming out. And I just wondered where 23 you guys were on it.
24 Personally -- and again, I'm sure Dave 25 will have us go around -- but I think you guys have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
221 1 done a great job. You've made good progress. But, 2 you know, interacting, we have some ideas. I think it 3 will be interesting to see what the public comments 4 are at the end of this meeting, as well as what you 5 learn in the middle of March, when you go back to them 6 and talk to them.
7 Anyway, that's just my thoughts.
8 CHAIRMAN PETTI: Other members?
9 MEMBER DIMITRIJEVIC: Yes, Dave. Hi.
10 This is Vesna.
11 I wanted to make one general comment in 12 the end about something which is very important for 13 me, sort of my hangup. And that was not addressed so 14 much in the source data or it was stressed in the 15 beginning of looking. And that's uncertainty 16 associated with results.
17 So, we saw some of your results where 18 there was a play with the parameters in Monte Carlo 19 and produced some uncertainties. But is just a 20 thought for an ice break. There is some much 21 uncertainties in these results which are associated 22 with models, you know, the radionuclides, barriers, 23 the facilities, the methodology which is used. So, 24 uncertainties are so much bigger than these parametric 25 uncertainties. And maybe there we see results NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
222 1 presented with uncertainties, the brackets.
2 So, I would like, actually, you know, and 3 this is also my personal opinion, but I think that 4 everybody agrees the uncertainties is a necessary part 5 of the game and something starting to be used in 6 regulation. You know, in my opinion, my personal 7 opinion, the uncertainties from the PRA Level 1, I 8 don't know if they necessarily are handled well, but 9 they are much better and astute, and they always 10 address and we talk about mean values and 95 and 11 median and point estimates.
12 However, as we progress through the Level 13 II and Level III, in my opinion, uncertainties triple.
14 They become a much bigger and much larger, and we 15 don't really see that yet, because these uncertainties 16 are not as well and astute, and they're not looking to 17 in the level of details necessary.
18 So, I understand that this is just in the 19 beginning and maybe it's too early to want a lot of 20 attention to uncertainty, but I definitely think that 21 that's something we should be very seriously looking 22 to.
23 All right. That was my comment. You 24 know, I don't think we really understand the 25 uncertainties associated with this and we need to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
223 1 devote much more attention to understand these 2 methods, models, and methodology associated with it.
3 CHAIRMAN PETTI: Other members?
4 Oh, go ahead, John. No, John, go ahead, 5 if you want to respond.
6 MR. SEGALA: Yes, I was just going to say 7 that I think we agree that uncertainties are 8 important. I think modeling and understanding the 9 uncertainties and accounting for the uncertainties are 10 something that's acknowledged in the Licensing 11 Modernization Project, as well as the non-light water 12 reactor PRA standard. I think that that is something 13 that is going to have to be accounted for in the 14 design margins, and whatnot, as we move forward.
15 So, thank you for the comment.
16 CHAIRMAN PETTI: Anyone else?
17 MEMBER KIRCHNER: Dave, this is Walt.
18 CHAIRMAN PETTI: Yes?
19 MEMBER KIRCHNER: I misquoted the number 20 of the Draft Reg. Guide earlier. I should have been 21 saying Draft Guide 1389. Is that the topic of your 22 Subcommittee's meeting in March? Are we going to 23 review that document?
24 CHAIRMAN PETTI: Yes, I think so. That's 25 -- we gave up on 1.138 one -- what it will evolve to, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
224 1 I believe.
2 MEMBER KIRCHNER: Yes. Well, just a 3 thought and observation. I want to go away and study 4 it more closely. But it seems that, echoing an 5 earlier observation, that the methodology used to 6 develop an alternate source term for an LWR, where we 7 have a lot more data, we have a lot more experience, 8 should, in general, and conceptually, apply for 9 advanced reactors.
10 And I'll take a look at the Reg. Guide 11 Draft, but it seems to me the LWR specifics could be 12 in appendices and the methodology or the recommended 13 approach, much like was covered earlier from the INL 14 report, could find its way into at least an executive 15 summary or an introduction to the Reg. Guide. I'll go 16 and take a look, and I'll give you some comments 17 perhaps that could be used for that interaction.
18 CHAIRMAN PETTI: Okay.
19 MEMBER KIRCHNER: And the other thing, 20 Vesna, I agree with you. Having had some experience 21 in this modeling and simulation business, once you 22 lose the fuel geometry, then the uncertainties in the 23 models are much greater than the parametric 24 uncertainties. You get into stochastic and random 25 kind of results.
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225 1 So, once you're in that territory, so to 2 speak, once the uncertainties expand, and bounding 3 analyses are probably a wise choice to try and get 4 some sense of the consequences -- this is true of even 5 thermal hydraulics for existing plants. Once you go 6 into two-phase flow and the phases separate, then the 7 uncertainty in the model results increases 8 substantially. And that has nothing to do with 9 parametric uncertainties. It's just the uncertainty 10 in terms of our ability to model what is happening 11 physically. Fuel fragmentation is a good example of 12 that.
13 MEMBER DIMITRIJEVIC: Yes. You know, 14 Walt, I was thinking, how can this be addressed? I 15 was thinking that, maybe since this is international, 16 there are other programs which do similar maybe 17 computing results, or in the case of Fukushima, we 18 have real data to compare with the results of the 19 models.
20 I don't know how to estimate that. It's 21 not an easy thing to do. I mean, even I don't think 22 the bounding things always is a way to address 23 something, especially as these become a more important 24 part in the regulation. I mean, how are we going to 25 do the risk-informed applications based on, you know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
226 1 source term, and the dose and dispersions, and things 2 like that?
3 So, I'm not sure how to address that, but 4 it's exactly what I thought, what you said. Thanks.
5 MEMBER KIRCHNER: And, Dave, I would just 6 like to thank the staff for this great presentation 7 today. There's a lot of information here.
8 CHAIRMAN PETTI: Yes.
9 MEMBER KIRCHNER: So, I look forward to 10 browsing or grazing on their website.
11 CHAIRMAN PETTI: Yes.
12 MEMBER KIRCHNER: Thank you.
13 CHAIRMAN PETTI: You took my comment. I 14 really want to reiterate what Walt just said. This 15 presentation, you know, the depth and breadth is 16 fairly unique from what we've heard, I think, in other 17 presentations. It required a lot of different folks 18 to come together, and I really appreciate it, given my 19 background, of course, in this area. I've thought 20 about these things as well, and I really do appreciate 21 the slides.
22 That said, we should discuss, among 23 ourselves I guess, the need for a letter. My letter 24 was going to be --
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227 1 first, though?
2 CHAIRMAN PETTI: You want comments first?
3 Okay. Sure.
4 MEMBER REMPE: Yes. Because I think that 5 I'd like to hear what the public has to say.
6 CHAIRMAN PETTI: Okay.
7 MEMBER REMPE: Is that okay?
8 MR. SNODDERLY: Yes, that's the way we had 9 -- we had it set up that way, before member 10 discussion.
11 CHAIRMAN PETTI: Okay. So, for members of 12 the public, *6 to unmute yourself. Give us your name 13 and your comment.
14 MR. SNODDERLY: I'm sorry. Dave, we did 15 something different today. So, the members of the 16 public are on. So, you just have to unmute yourself.
17 CHAIRMAN PETTI: Oh.
18 MR. SNODDERLY: Yes. Yes.
19 CHAIRMAN PETTI: Okay.
20 MR. SNODDERLY: I don't know if Ms. Fields 21 is still there, but I saw some members of the public 22 in.
23 MS. FIELDS: Hi. Yes, this is Ms. Fields.
24 Can you hear me?
25 CHAIRMAN PETTI: Yes, we can.
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228 1 MS. FIELDS: Oh, okay. Because of another 2 appointment, I didn't hear the whole presentation, but 3 what I did hear and see was very, very informative.
4 So, I just have a few comments.
5 As you know, there's still no permanent 6 repository for irradiated nuclear fuel. And to me, it 7 does not make sense for the United States to keep 8 making used or spent nuclear fuel when there is no 9 permanent repository and none anticipated in the near 10 future, and I believe none in my lifetime.
11 Also, the NRC should not use the term 12 "advanced" in these various rulemakings and documents.
13 "Advanced" is a public relations term. It doesn't 14 have a regulatory or a statutory or a technical basis.
15 The NRC has already dropped the term 16 "advanced" in the Part 53 rulemaking. The rulemaking 17 is not for advanced nuclear reactors; it's just for 18 commercial nuclear reactors. And the NRC has taken 19 the definition of "advanced" out of that rulemaking.
20 And there are some various reasons for 21 that -- in part, I imagine, because the NEIMA 22 definition of "advanced" just does not make sense.
23 There's just no real basis for that.
24 Also, I'm still concerned that the NRC has 25 not been able to accession properly the documents NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
229 1 related to these pre-application and application 2 submittals to the NRC. Many of the documents have 3 been accessioned to ADAMS, but they have no docket 4 number, or the docket number should be on an 5 application document for an application docket, but 6 it's still on the pre-application docket. The NRC has 7 been paying attention to this and made some changes, 8 but there's still a big issue.
9 I also think that the NRC and the ACRS 10 should pay more attention to the needs of the public.
11 Obviously, this whole process and the discussion today 12 was really to satisfy the needs of the industry, as 13 they move forward with these new reactor designs and 14 proposals.
15 But there's little information available 16 to the public that would give them an idea of what 17 exactly they should look for, if there is an 18 application to site one of these new reactor designs 19 in their community. There's just really not much 20 there that would guide the public, as they review 21 these applications and they consider whether this is 22 appropriate for their community.
23 So, I definitely think that the NRC should 24 think -- well, maybe even have a public meeting for 25 public input on what exactly the public, information NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
230 1 the public needs, not so that the industry can site 2 something in their community, but so the community has 3 the tools it has to actually evaluate what's going on, 4 like the amount of water that might be needed. Or is 5 it really appropriate to reduce the EPZ and not have 6 it at the traditional 10 miles?
7 So, it's just something for you all to 8 think about in the future. Thank you.
9 CHAIRMAN PETTI: Thank you.
10 I believe Kati Austgen is next.
11 MS. AUSTGEN: Yes, thank you. Kati 12 Austgen with the Nuclear Energy Institute.
13 First, I just wanted to thank the ACRS and 14 the NRC staff for trying something new and letting us 15 participate in the Teams webcast this time around. It 16 was much easier to follow along on the slides and 17 understand which members and which staff were making 18 which comments. So, we very much appreciate that.
19 And then, I agree with the ACRS that this 20 was a very informative presentation, and I look 21 forward to sharing the information about the new 22 website with our members, and continuing our 23 conversations with the staff on guidance in this area 24 and how we can all make sure we're on the same page 25 about what's required.
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231 1 Thank you.
2 CHAIRMAN PETTI: Thank you.
3 Jan Boudart?
4 MS. BOUDART: Jan Boudart, and I'm a Board 5 member of the Nuclear Energy Information Service in 6 Chicago.
7 And apropos to what Sarah Fields said, 8 there is being considered an experimental reactor at 9 the University of Illinois in Champaign-Urbana. And 10 the question that NEIS is asking is, how do people 11 feel about sending their kids to that University if 12 there's going to be a fission project on the 13 University campus? And how is the NRC, or whatever 14 appropriate agency -- maybe it's the DOE -- how are 15 they going to present this to the public and to the 16 people who are considering sending their kids there?
17 And is there going to be a public hearing? And will 18 the ramifications of having this on the campus be 19 explored, et cetera, et cetera?
20 But I think this is very appropriate to 21 what Sarah said, because the people who are going to 22 be affected by these small modular nuclear reactors 23 and the experimental reactors, like the one that's 24 being considered with the TVA in Tennessee, the people 25 in the neighborhood and the people who are going to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
232 1 in the area need to be consulted as to whether they 2 think it's appropriate to have these things near them.
3 That's the end of my comment.
4 CHAIRMAN PETTI: Thank you.
5 Any other comments from members of the 6 public?
7 MEMBER REMPE: Dave, am I -- oh, I'm 8 sorry, I interrupted. Yes.
9 CHAIRMAN PETTI: Please go ahead.
10 MR. STEIN: This is Adam Stein from the 11 Breakthrough Institute. I have a comment and a 12 clarifying question, if you are willing to entertain 13 the question.
14 To what extent does the staff expect to 15 require a MELCOR model to be developed for specific 16 advanced reactors? Because, as was presented today, 17 the models that have been created so far are not 18 direct replicas of the various developers; they are 19 based on other similar technologies, or similar 20 designs, I should say. But, as the staff mentioned 21 today, a significant amount of time and effort was put 22 into developing those models. So, to what extent is 23 the staff expecting to require a MELCOR model that is 24 similar to an applicant's design to be developed? And 25 if not, how closely resembling results are they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
233 1 expecting between the existing MELCOR models or the 2 ones that are being developed and an applicant's 3 submission of results?
4 CHAIRMAN PETTI: Thank you. We don't 5 respond to comments from the public in our meeting.
6 But you can send an email to Mike Snodderly, the 7 Designated Federal Official. Mike can tell you his 8 email, and take it from there.
9 MR. SNODDERLY: Yes, Adam, I think if you 10 want to send me a written comment, then we can add 11 that to the record. But the other option is you just 12 turn your question into a comment, and that also is 13 being transcribed.
14 So, what I heard could be your comment is 15 that MELCOR models, as they've been developed, are 16 valuable and should be required or part of any review 17 of a confirmatory analysis of a review of advanced 18 reactors.
19 But you just need to form it in that kind 20 of a phrasing. And if you want more time to do that, 21 you can do that, and then, send me an email to 22 michael.snodderly -- S-N-O-D-D-E-R-L-Y -- @nrc.gov.
23 MR. STEIN: Thank you.
24 I would also like to comment that it is 25 important to consider that, when you are looking at NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
234 1 the differences between a MELCOR model and a, 2 potentially, applicant-developed model, that it is 3 important to consider and disposition why and how they 4 may differ in results. Because, as was discussed 5 today, there are still various considerations or 6 physical parameters that are not in both models 7 identically, and there will still be dispositions.
8 Thank you for your time.
9 CHAIRMAN PETTI: Any other public 10 comments?
11 I see a hand. Jan?
12 MS. BOUDART: It's me again.
13 I wanted to ascertain whether Mr.
14 Snodderly's email, does it say @acrs.nrc.gov or is it 15 just nrc.gov?
16 MR. SNODDERLY: Yes, Jan, it's @nrc.gov.
17 MS. BOUDART: Oh, good. Oh, thank you.
18 That's what I thought. Thanks.
19 MR. SNODDERLY: Thanks.
20 CHAIRMAN PETTI: Okay.
21 MEMBER REMPE: Mike, is your hand up?
22 MR. SNODDERLY: Yes. If you don't mind, 23 Dave, I just wanted to remind the Committee, as you 24 start your member discussion, I went back to your 25 presentation on October 8th to the Commission. And in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
235 1 that, just to remind everybody, in your slides that 2 you presented -- let's see -- the second-to-last slide 3 that you presented, you said that, "Numerous recent 4 and upcoming source-term-related activities. A 5 roadmap showing how all the pieces fit together would 6 be worthwhile"; and that "Many different pieces are 7 coming together, and that the ACRS plans an integrated 8 review later this year."
9 So, I just wanted to remind the Committee 10 that, well, you didn't make a commitment, but you told 11 them that you were going to have this interaction.
12 You've had it. It doesn't require a letter, of 13 course. But I think the staff has done something to 14 provide somewhat of a roadmap, and you may want to 15 comment on how good or bad it is. But you, of course, 16 can say nothing. But I just thought that might be a 17 good kickoff point, and now, I'll be quiet.
18 MEMBER REMPE: Mike, also, while you're on 19 the line, just so I understand the new process, 20 everyone can just -- does anyone need to press *6 now?
21 Because there are a lot of phone lines out there.
22 MR. SNODDERLY: No, they're on just like 23 you and I, and it's mute and unmute.
24 MEMBER REMPE: Okay.
25 MR. SNODDERLY: Now I can mute folks, any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
236 1 participate, but --
2 MEMBER REMPE: Right, but the phone line 3 folks --
4 MR. SNODDERLY: Yes.
5 MEMBER REMPE: I understand the folks that 6 are linked in. Like Sarah has a real link.
7 MR. SNODDERLY: Yes.
8 MEMBER REMPE: But the phone line folks 9 can it themselves? Okay. Just wanted to make sure we 10 weren't leaving anyone out. Thank you.
11 MR. SNODDERLY: That's my understanding, 12 but Tom Daschle (phonetic) -- you know, correct me if 13 I'm wrong -- but I think that --
14 MR. BURKHART: This is Larry.
15 One of the commenters, our last commenter 16 was on the phone. So, I think it's working.
17 MR. SNODDERLY: Yes. Thank you, Larry.
18 MEMBER REMPE: Great. Thank you.
19 CHAIRMAN PETTI: Okay. So, Members, let's 20 talk a little bit about the need for a letter; what 21 your thoughts are.
22 I have, as I always have, put together 23 what I call the guts of a letter. It's just a bunch 24 of points and starting points.
25 While I don't think the staff needs a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
237 1 letter, my opinion is that, in light of everything 2 that's going on right now in the area with non-LWRs, 3 that a pointer letter, which is what the letter would 4 be, pointing out some of the important things that 5 they've done, might be quite helpful. And I'm 6 certainly willing to write the letter, given the 7 importance of the topic, as we heard from (audio 8 interference) for the NRC.
9 So, comments, Members? And consultants?
10 I'd be interested in hearing their perspectives, too.
11 MEMBER BROWN: It's Charlie.
12 Based on my listening, this was kind of a 13 potpourri-type presentation. I head a lot of comments 14 from us, but unless we have some specific things that 15 should go on in this general, multiple opportunities 16 for people, I don't see where a letter would really 17 add a whole lot, other than just saying, "Continue."
18 My opinion is I wouldn't write a letter based on this 19 meeting.
20 MEMBER KIRCHNER: Dave, this is Walt.
21 I guess I would concur with Charlie. I 22 would wait until you heard, we hear about DG-1389 and 23 its status. Because what I'm thinking is, although 24 this set of presentations today primarily is 25 addressing, quote-unquote, "advanced" reactors and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
238 1 applicants coming in, it's likely that they're going 2 to go in the near term, while 10 CFR 53 rulemaking is 3 in progress, that they are going to come in under 4 10 CFR 50 or 52. And if they do so, then they can 5 either do exceptions to guidance and the regulations 6 or they'll follow the spirit of the guidance that's 7 out there. And that's why I'm very interested in what 8 DG-1389 contains and whether it's a broader template 9 for other advanced reactor applications.
10 Thank you.
11 CHAIRMAN PETTI: Go ahead, Ron.
12 MEMBER BALLINGER: This is Ron.
13 I think that, unless we want to reinforce 14 something that we've already communicated on at least 15 two occasions -- the first with the Commission 16 presentation and the second, at least the second today 17 -- I think that we should probably -- I think the 18 staff has got the message. And so, I think I'm in 19 agreement with Walt -- waiting until we have something 20 which we can actually review in detail.
21 Thank you.
22 CHAIRMAN PETTI: My only concern, having 23 read that Draft Guide, is that it's extremely light-24 water-reactor-specific and we're going to focus on 25 that. So, I don't see us getting into non-LWRs in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
239 1 that review because that's not where that is.
2 MEMBER BALLINGER: That would not keep us 3 from making a comment, does it?
4 MEMBER KIRCHNER: That's what I was 5 thinking, Dave. You can build on that to say whatever 6 we feel -- more, less, or keep on going, or whatever 7 -- for the non-LWR reactors or something new, 8 completely new and different is needed. That would be 9 a good juncture to make that kind of observation.
10 MEMBER HALNON: Dave, this is Greg.
11 CHAIRMAN PETTI: Dennis? Oh, go ahead, 12 Greg.
13 MEMBER HALNON: No, well, Dennis had his 14 hand up. I couldn't see that.
15 Go ahead, Dennis. I'll come after you.
16 DR. BLEY: Okay. Yes, Dave, well, first, 17 I'd say I found today very useful, and I've played 18 around on that website already this afternoon. I 19 think that's going to be especially nice.
20 Now we sometimes get tricked in things we 21 read, as Joy said, in the popular press and in papers 22 people send to us to look at. I'm not sure how much 23 of what I read is politics and how much is real, where 24 people are technically.
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240 1 that kind of say, "Gee, the NRC is stuck in the Dark 2 Ages and they ought to just say, almost, you don't 3 need to do anything with these new reactors. Just let 4 them operate."
5 I think what we saw today ties a lot of 6 very important information together. And if I had a 7 new design and it had extremely low consequences and 8 likelihood of damage, I think I could do something not 9 terribly elaborate to prove that point within those 10 constructs.
11 So, I think what they've put together 12 today is really useful, integrating all these pieces.
13 I would have agreed with Walt and Ron that we ought to 14 wait for the meeting next month, and there's when it 15 would be most useful for you folks to write a letter.
16 But I didn't read the draft and you have. And if they 17 aren't closely coupled, then, maybe it makes sense 18 separately.
19 I mean, this isn't at a point where this 20 rates a 10-page letter. A one-page letter might be 21 appropriate to acknowledge what's being done and its 22 value. So, I think we shouldn't let this pass without 23 -- I don't think you should let it pass without some 24 comment.
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241 1 points are all of 80 lines. You know, it's not a big 2 letter. But we're viewed as sort of an independent 3 look at things, and there's these questions out there 4 by all sorts of folks about NRC, as you say, being in 5 the Dark Ages and they don't really understand these 6 non-LWRs. And I think nothing could be further from 7 the truth. They've done an awful lot of work to get 8 themselves ready to accept an application. And to 9 recognize that, I thought, personally, was of value.
10 MEMBER HALNON: This is Greg.
11 I look at it from a perspective of the 12 Commissioner sitting in his or her office, and 13 somebody walking in wanting a drop-in meeting and have 14 discussion about some iterations they're going through 15 with reviews, and whatnot.
16 And I guess there's a couple of nuggets 17 that came out today. And notwithstanding the one 18 about the burden of developing the methodologies is on 19 the developer and expect it to be complicated, and 20 those sorts of things, there's a couple of messages 21 that I think the Commissioners need to hear, so that 22 they're armed with some information.
23 And I'm not sure if it gets to them or 24 not, but the fact that there is an integrated website 25 and it's in process; it's in really pretty good shape, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
242 1 but we're taking comments on it; the fact that there's 2 a couple of nuggets in there that we heard today that 3 we're either expecting or anticipating or have learned 4 -- I think those type of things, going to the 5 transcripts, might be useful.
6 And I'm not talking about a long letter, 7 but a couple of sentences here and there that 8 acknowledge that we're in an iterative-plus process 9 that's going to be getting better every time we have 10 interaction with a new technology and we all learn 11 more about it. I think those types of things are 12 pretty important.
13 So, I'm kind of in with you and Dennis, 14 Dave, that there could be a very short letter that has 15 some of the things that we heard, some of the things 16 that we anticipate, and wouldn't negate having another 17 letter down the road for a specific Reg. Guide, but, 18 certainly, from the topic itself, could use some, I 19 guess, collation into a concise letter.
20 CHAIRMAN PETTI: Steve? I see your hand 21 up.
22 DR. SCHULTZ: Yes. For the Committee, I 23 do appreciate the discussion here about a letter, but 24 I don't think it's letter or no letter. Because I 25 think the Committee has made these comments to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
243 1 Commission in their last meeting; that this is a topic 2 that is extremely important to the advancement of the 3 new reactors; and also, to the work that is being done 4 associated with their development and application.
5 And I think a followup to just those 6 comments to the Commission would be very useful, 7 especially in light of the excellent presentations 8 that were made today. A very well-integrated 9 presentation and very useful to the Committee and, 10 also, in the public forum.
11 My other comment would be that not only is 12 this of interest in the United States, but the 13 international community. The regulatory agencies in 14 those areas are also working on the same thing. And 15 for the ACRS to not come out with a continuing 16 statement about what is ongoing and what is 17 appreciated and associated with this work -- I think 18 that's important for the Committee to do. Whether 19 it's one letter or two letters, it's up to the 20 Committee. I think that an individual letter on this 21 topic would be not the last letter the Committee 22 writes, but very important to continue to endorse.
23 CHAIRMAN PETTI: Joy?
24 MEMBER REMPE: So, I'm torn. Clearly, we 25 need to decide because the month is coming up and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
244 1 staff needs to know if they're going to be coming back 2 to us.
3 But, listening today to the public 4 comments, it seemed to me, and what the staff said, 5 that they're going to have an upcoming stakeholder 6 meeting in March, where they'll maybe learn more about 7 whether more guidance is needed -- I mean, clearly, 8 they've made a lot of good progress and they've got a 9 good -- you know, I really do like the reference plant 10 evaluations. I'd like to see a little more coming out 11 of it, which I think the staff was receptive to.
12 I think that some of our comments about 13 the need for guidance is important, but I'm not sure 14 whether we should be telling the staff to do that, 15 when I really appreciate that there's so many 16 different designs out there, that maybe the staff is 17 at the right -- that what they're doing is correct.
18 Because why waste resources on doing guidance for so 19 many different designs and flexibilities?
20 And one of the public comments, to me, 21 indicated that the person didn't understand the 22 staff's broad perspective. I don't think the staff is 23 trying to say, by any means, that a MELCOR model is 24 needed in what we've heard today. So, I think 25 communication with the stakeholders is very important.
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245 1 But that was my impression of it and all, 2 but I don't know if the timing is right. I definitely 3 would put "Interim Letter on Staff Progress," or 4 something like that, as the title of whatever is 5 issued. But I'm open to whatever the Committee 6 decides, but I don't know if this is the right time or 7 not.
8 I'm not sure that's a very helpful 9 comment.
10 CHAIRMAN PETTI: My biggest concern is, 11 you know, there's probably never a good time. And we 12 know the bow wave of what's coming in front of us, 13 starting in April. There just seemed to be a window 14 of opportunity here, given what's on the plate for 15 March.
16 MEMBER REMPE: But what will we say other 17 than, "Good job."? And we can do that in the minutes 18 of the meeting summary, when you come back and do 19 something like what Jose often does. I don't know; I 20 just am wondering if we're making them come back to 21 present to us and things like that. So, anyway, I'm 22 just kind of thinking about it.
23 CHAIRMAN PETTI: Anybody else? Matt?
24 Jose?
25 MEMBER MARCH-LEUBA: Yes, Jose. Jose has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
246 1 his hand raised because Joy used my name in vain, and 2 she took my argument out of it.
3 We have an intermediate method of dealing 4 with this. First, let me say that I'm ambivalent on 5 the letter. On the one side, the person that knows 6 and cares more about this is Dave, and he thinks we 7 should have a letter. And that makes me want to 8 support it. On the other side, I see all the other 9 arguments from the people that maybe we should wait 10 for a real Reg Guide.
11 But we have a middle ground, which is 12 writing a couple of paragraphs to be included in the 13 summary of this meeting. And they get presented to 14 P&P, and they become the property of ACRS. It's not 15 a subcommittee any longer. I mean, once P&P approves 16 those two or three paragraphs, it's half a letter, and 17 those paragraphs can be submitted by email to 18 whomever.
19 So, if we are voting, I will vote for the 20 halfway and go for the summary.
21 MEMBER REMPE: And that avoids the staff 22 having to come back and do another presentation.
23 CHAIRMAN PETTI: Matt, you've been silent.
24 MEMBER SUNSERI: I've been listening, 25 Dave. You know, I think it was a very important NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
247 1 topic. We've been asking for how these various source 2 term activities integrate each other towards the 3 topic. And so, I'd be in favor of a letter.
4 MEMBER BIER: This is Vicki.
5 It's kind of out of my area. I feel sort 6 of similar, in between Charlie and Dennis; that, you 7 know, on the one hand, there was nothing burning today 8 that inspired me that we have to write a letter, but 9 I could certainly go with a short letter that just 10 said, "We support the progress being made," et cetera.
11 CHAIRMAN PETTI: Okay.
12 MEMBER DIMITRIJEVIC: Dave?
13 CHAIRMAN PETTI: Yes?
14 MEMBER DIMITRIJEVIC: You know, I believe 15 in Jose's philosophy. You know most about that and 16 you put the letter together, and you think it should 17 be written. So, I support that.
18 MEMBER SUNSERI: Dave, this is Matt. One 19 more thing.
20 I think that we have precedents, also, for 21 not having to have staff come back and make another 22 presentation to the Committee. I think maybe almost 23 everybody from the Committee is here, and that you 24 could make some kind of presentation, a short one, at 25 the full Committee meeting, and we could write the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
248 1 letter from that. I think we've done that before on 2 a lot of the license reviews that we did for design 3 certifications.
4 MEMBER REMPE: We did it for MELLA+,
5 actually, and that would make it easier for the staff.
6 I don't know.
7 CHAIRMAN PETTI: Yes. I certainly don't 8 -- I mean, we would need a very, very, 100,000-foot 9 level presentation, because this here, I'm thinking I 10 would have thought no more than five slides sort of 11 thing. You know, what were the things you touched on?
12 But if we can even get rid of that, maybe, then, I 13 could put a draft together, and then, when we see it, 14 that might even inform us better. To bring it to the 15 full Committee, and then, people make a decision? Is 16 that allowable?
17 MR. SNODDERLY: Yes, I believe that's 18 consistent with the Bylaws. The Subcommittee needs to 19 make a recommendation to the full Committee on whether 20 to write a letter or not. This would also give some 21 more time to digest it and think about it.
22 But, Dave, you would, when we come to this 23 item on the agenda, you would make, as a member, a 24 recommendation to the Committee, and then, the 25 Committee would vote whether to take it up further and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
249 1 write the letter or not, or to do some hybrid thing.
2 And it's already been discussed or the 3 precedent has been set. The staff could be there or 4 not be there. But, of course, we would like them 5 there to support letter writing as usual, anyway. But 6 I don't even know what, if any, presentation would be 7 needed, unless you requested it, what you think you 8 need to support the letter.
9 CHAIRMAN PETTI: All right. I don't think 10 we need them to be there to make a presentation.
11 But let me try to write --
12 MEMBER REMPE: Let me interrupt for a 13 minute, and I think Larry is going to have to weigh-in 14 because the agenda is already published in The Federal 15 Register.
16 If we were to do this in P&P, that puts 17 this on Friday morning, just to be thinking 18 processwise here and the mechanics of it. I don't 19 think we can have -- or maybe we can; I'm asking 20 Larry. Is it acceptable for Dave to give a 21 presentation during that allocated time for this 22 topic, Larry, instead of the staff, and go ahead with, 23 if the Committee at that point decides to do a letter, 24 is it acceptable to go ahead and go right into letter 25 writing?
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250 1 MR. BURKHART: I think it's perfectly 2 acceptable for the Chair of the Subcommittee to orient 3 the full Committee on the topic, and then, move into 4 report writing, if that is decided.
5 MEMBER REMPE: Okay. I just wanted to 6 make sure, because this is a little different than 7 what happened with the MELLA+ thing. But that's 8 great. I just wanted to make sure we weren't going to 9 go forward with something that wasn't allowed. Thank 10 you.
11 CHAIRMAN PETTI: The other thing is what 12 Mike raised. You know, we did talk about this with 13 the Commissioners, and it would be nice to kind of 14 close the loop on that. Again, a shorter letter, that 15 both sides recognize this issue; thought it was 16 timely, and we did.
17 So, it's on the agenda. I will, 18 basically, I guess, during that slot, talk about the 19 major points in the letter. And we can have a 20 discussion there, and then, make a decision, as the 21 full Committee, whether we want a document for the 22 record or, in a summary sense, in the meeting summary.
23 Is that reasonable?
24 MEMBER SUNSERI: Sounds good to me.
25 MEMBER DIMITRIJEVIC: Absolutely.
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251 1 Absolutely.
2 CHAIRMAN PETTI: Okay. I will work on 3 this next week, and I'll actually try to get some of 4 the stuff to folks before the meeting, so they're not 5 seeing for the first time at full Committee.
6 Okay. Then, with that, I guess if there 7 are no other comments, we're ready to adjourn the 8 meeting.
9 And I thank everyone and, again, thank the 10 staff. This was a Herculean effort. We really do 11 appreciate it.
12 So, everybody have a good evening. Thank 13 you.
14 (Whereupon, at 4:28 p.m., the meeting was 15 adjourned.)
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Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives February 17, 2022 1
AGENDA
- Opening Remarks
- Staff Introduction
- History and Evolution of LWR Source Term
- NRC analytical tools and past studies
- SCALE/MELCOR non-LWR reference plant analysis Break
- Agenda Item IV Continued
- NuScale EPZ Sizing Methodology Topical Report, Rev. 2
- Light water SMR design certification source term approach
- Source term approach for early non-LWR movers Lunch
- Accident-consequence-related regulation activities Break
- Guidance and information for developing advanced reactor source term
- Guidance for developing advanced reactor source term (long-term)
- Opportunity for Public Comment
- Member Discussion Adjourn ACRS meeting on Integration of Source Term Activities in Support of Advanced 2
Reactor Initiatives, 02/17/2022
Integration of Source Term Activities in Support of Advance Reactor Initiatives John Segala NRR/DANU February 17, 2022 3
Staff Introduction
- Determining source terms is a critical component in the NRCs licensing process
- NRC team presenting today:
- Mark Blumberg - NRR/DRA
- Michelle Hart - NRR/DANU
- Jason Schaperow - NRR/DANU
- Bill Reckley - NRR/DANU
- Tim Drzewiecki - NRR/DANU
- Hossein Esmaili - RES/DSA ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 4
Reactor Initiatives, 02/17/2022
Integration of Source Term Activities in Support of Advance Reactor Initiatives John Segala NRR/DANU February 17, 2022 5
ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 6
Reactor Initiatives, 02/17/2022
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NRC Analytical Tools and Past Studies-Severe Accident Progression and Source Term Hossein Esmaili, RES/DSA Jason Schaperow, NRR/DANU 16
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https://www.nrc.gov/reactors/new-reactors/advanced/details.html#non-lwr-ana-code-dev ACRS meeting on Integration of Source Term Activities in Support of Advanced 46 62 Reactor Initiatives, 02/17/2022
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NuScale EPZ Sizing Methodology Topical Report, Rev. 2 Light Water SMR Design Certification Source Term Approach Source Term Approach for Early non-LWR Movers ACRS meeting on Integration of Source Term Activities in Support of Advanced 79 Reactor Initiatives, 02/17/2022
Accident Source Term in Recent and Near-term Applications Michelle Hart NRR/DANU/UTB2 80
Outline
- SMR and non-LWR accident source terms recent experience
- Emergency planning zone size justification consequence analyses
- Example: SMR design certification source term approach
- Source term approaches for non-LWR early movers ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 81 Reactor Initiatives, 02/17/2022
SMR and Non-LWR Accident Source Terms Recent Experience
- Advanced reactor pre-application interactions, topical report reviews, and license applications
- Source term development contractor reports ACRS meeting on Integration of Source Term Activities in Support of Advanced 3 82 Reactor Initiatives, 02/17/2022
Emergency Planning Zone Size Justification Consequence Analyses
- Concept based on NUREG-0396
- Technical basis for plume exposure and ingestion pathway EPZ radius of ~10 and ~50 miles, respectively
- Identification of area within which prompt protective actions may be necessary to provide dose savings in the event of a radiological release
- Calculate dose at distance for a spectrum of accidents
- Analysis includes design basis accidents and severe accidents ACRS meeting on Integration of Source Term Activities in Support of Advanced 4 83 Reactor Initiatives, 02/17/2022
Emergency Planning Zone Size Justification Consequence Analyses
- No separate/unique source terms developed especially for EPZ size analysis
- Re-use source terms and accident release information developed for safety analysis report and PRA ACRS meeting on Integration of Source Term Activities in Support of Advanced 5 84 Reactor Initiatives, 02/17/2022
Emergency Planning Zone Size Justification Consequence Analyses
- Methodology to support exemptions to 10-mile requirement
- Clinch River ESP EPZ size methodology described in SSAR
- Methodology to support plume exposure pathway EPZ size determination on case-by-case basis for reactors <250 MWt
- NuScale EPZ sizing methodology topical report (under review)
- SECY-22-0001 issued for Commission review and approval
- Guidance on analysis in appendices to RG 1.242 ACRS meeting on Integration of Source Term Activities in Support of Advanced 6 85 Reactor Initiatives, 02/17/2022
NuScale EPZ Sizing Methodology Topical Report
- TR-0915-17772, Revision 2, submitted in 2020, currently under review
- Not part of DC review
- Applicable to light-water SMRs such as NuScale, although not limited to the NuScale designs
- Rev. 3 under development
- Analysis methodology to determine plume exposure pathway EPZ size ACRS meeting on Integration of Source Term Activities in Support of Advanced 7 86 Reactor Initiatives, 02/17/2022
NuScale EPZ Sizing Methodology Topical Report
- Source term refers to fission product release to the environment as a function of time
- No separate/unique source terms developed especially for EPZ size analysis
- Uses CDF from PRA to categorize severe accidents and select accident sequences to evaluate against relevant dose criteria ACRS meeting on Integration of Source Term Activities in Support of Advanced 8 87 Reactor Initiatives, 02/17/2022
Example: SMR Design Certification Source Term Approach
- SECY-19-0079, August 16, 2019
- Describes staff review approach to evaluate accident source terms for both the TR and the NuScale SMR DC application
- Provides basis for using source term without core damage for environmental qualification ACRS meeting on Integration of Source Term Activities in Support of Advanced 9 88 Reactor Initiatives, 02/17/2022
Example: SMR Design Certification Source Term Approach - NuScale TR
- NuScale TR-0915-17565, Accident Source Term Methodology, Revision 4, February 2020
- Methods to develop accident source terms are consistent with RG 1.183 guidance for PWRs except for:
- Core damage source term for Core Damage Event
- Iodine spike design basis source term (no fuel damage)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 10 89 Reactor Initiatives, 02/17/2022
NuScale TR: Core Damage Event
- Derive source term from range of accident scenarios that result in significant damage to the core
- NuScale-design-specific analyses using MELCOR to be performed by applicant referencing the TR
- Radionuclide transport phenomena
- Iodine retention in containment based on pH
- Aerosol natural deposition in containment ACRS meeting on Integration of Source Term Activities in Support of Advanced 11 90 Reactor Initiatives, 02/17/2022
NuScale SMR DC Application: Core Damage Event
- Implemented the NuScale TR methodology to determine the core damage source term
- Core inventory calculated using SCALE code
- Scenario selection
- Based on NuScale SMR PRA, internal events
- 5 surrogate scenarios
- Various failures of ECCS, with decay heat removal system available
- Intact containment ACRS meeting on Integration of Source Term Activities in Support of Advanced 12 91 Reactor Initiatives, 02/17/2022
NuScale SMR DC Application: Core Damage Event
- MELCOR used to estimate release timing and magnitude for each scenario
- Release onset and duration from scenario with minimum time to core damage
- Core release fractions taken as median of scenarios
- Time-dependent aerosol removal rates calculated using STARNAUA code
- Design-specific input thermal hydraulic conditions calculated by MELCOR for surrogate scenario with minimum time to core damage ACRS meeting on Integration of Source Term Activities in Support of Advanced 13 92 Reactor Initiatives, 02/17/2022
Source Term Approaches for Non-LWR Early Movers
- Kairos Power
- MST methodology TR (under review)
- Methodology for applicants to develop event-specific radiological source terms
- DBAs for siting and safety analysis
- Hermes CP application (under review)
- Evaluates MHA, deterministic
- Refers to MST TR ACRS meeting on Integration of Source Term Activities in Support of Advanced 14 93 Reactor Initiatives, 02/17/2022
Source Term Approaches for Non-LWR Early Movers
- X-energy
- Proposed to use developer-made source term code (XSTERM) which includes modeling of radionuclides from generation to release (and dose)
- TR was submitted, but withdrawn to clarify and resubmit in future (not currently under review)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 15 94 Reactor Initiatives, 02/17/2022
Source Term Approaches for Non-LWR Early Movers
- Oklo Aurora COL application (review ended)
- Proposed maximum credible accident without release
- TerraPower
- Development of source term methodology described in 1/13/2022 public meeting (ML22011A072)
- Topical report planned for April 2023
- Terrestrial, Westinghouse, Others
- Source terms to be determined
- Public website information on non-LWR pre-application activities ACRS meeting on Integration of Source Term Activities in Support of Advanced 16 95 Reactor Initiatives, 02/17/2022
Acronyms AOO anticipated operational occurrence CDF core damage frequency COL combined license CP construction permit DBA design basis accident DBE design basis event DC design certification ECCS emergency core cooling system EP emergency preparedness EPZ emergency planning zone ESP early site permit FSAR final safety analysis report LMP Licensing Modernization Project MHA maximum hypothetical accident MST mechanistic source term MWt megawatts thermal Non-LWR non-light water reactor ONTs other new technologies PRA probabilistic risk assessment PWR pressurized water reactor RG regulatory guide SMR small modular reactor SSAR site safety analysis report TR topical report ACRS meeting on Integration of Source Term Activities in Support of Advanced 17 96 Reactor Initiatives, 02/17/2022
LUNCH ACRS meeting on Integration of Source Term Activities in Support of Advanced 97 Reactor Initiatives, 02/17/2022
Accident Consequence-Related Regulation Activities Michelle Hart NRR/DANU/UTB2 98
Petition for Rulemaking
- PRM-50-121, Voluntary Adoption of Revised Design Basis Accident Dose Criteria
- Received 11/23/2019, docketed 2/19/2020 (85 FR 31709)
- Under evaluation - no disposition yet
- Requests voluntary rule to allow power reactor licensees to adopt alternative to the accident dose criteria specified in § 50.67, Accident source term.
- Proposes a uniform value of 100 milli-Sieverts (10 rem) for offsite locations and for the control room ACRS meeting on Integration of Source Term Activities in Support of Advanced 19 99 Reactor Initiatives, 02/17/2022
Emergency Preparedness for SMRs and Other New Technologies Rulemaking
- Final rule in development
- New section 10 CFR 50.160, and related/conforming changes
- ACRS meetings in September and November 2021
- RG 1.242 (to be issued with final rule)
- Appendices
- Generalized analysis methodology
- Information on source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced 20 100 Reactor Initiatives, 02/17/2022
Emergency Preparedness for SMRs and Other New Technologies Rulemaking
- Appendix A, General Methodology for Establishing Plume Exposure Pathway Emergency Planning Zone Size
- Provides general guidance on the consequence analysis to support plume exposure pathway EPZ size determination
- Discusses event selection and consideration of accident likelihood ACRS meeting on Integration of Source Term Activities in Support of Advanced 21 101 Reactor Initiatives, 02/17/2022
Emergency Preparedness for SMRs and Other New Technologies Rulemaking
- Appendix B, Development of Information on Source Terms
- Provides guidance to develop source terms for plume exposure pathway EPZ size evaluations ACRS meeting on Integration of Source Term Activities in Support of Advanced 22 102 Reactor Initiatives, 02/17/2022
Alternative Physical Security for Advanced Reactors Rulemaking
- Draft rule and guidance in development
- Voluntary alternative physical security requirements commensurate with potential safety and security consequences
- Analyses (guidance under development)
- Develop relevant scenarios
- Site-specific potential offsite radiological consequences ACRS meeting on Integration of Source Term Activities in Support of Advanced 23 103 Reactor Initiatives, 02/17/2022
Acronyms CFR Code of Federal Regulations EPZ emergency planning zone FR Federal Register PRM petition for rulemaking RG Regulatory Guide SMR small modular reactor ACRS meeting on Integration of Source Term Activities in Support of Advanced 24 104 Reactor Initiatives, 02/17/2022
Guidance and Information for Developing Source Terms for Non-LWRs Michelle Hart, NRR/DANU/UTB2 Bill Reckley, NRR/DANU/UARP Tim Drzewiecki, NRR/DANU/UTB1 105
Outline
- Accident consequence analysis for advanced reactors
- Mechanistic source term
- Recent reports on Non-LWR source term development
- Non-LWR PRA standard and source term
- Licensing Modernization Project and source term
- Overview of method in NUREG-2246, Fuel Qualification for Advanced Reactors
- Non-LWR accident source term information website ACRS meeting on Integration of Source Term Activities in Support of Advanced 26 106 Reactor Initiatives, 02/17/2022
Accident Consequence Analysis for Advanced Reactors
- Regulatory nexus
- Siting and safety analysis regulatory requirement
- Newer uses for advanced reactors
- LMP
- Plume exposure pathway EPZ size determination
- Alternative security requirements - ongoing rulemaking
- Part 53 - ongoing rulemaking ACRS meeting on Integration of Source Term Activities in Support of Advanced 27 107 Reactor Initiatives, 02/17/2022
Accident Consequence Analysis for Advanced Reactors
- Accident source term development considerations
- Event selection, scenarios
- Balance of prevention vs. mitigation
- Relationship to functional containment
- A barrier, or set of barriers taken together, that effectively limit the physical transport of radioactive material to the environment (SECY-18-0096)
- Relationship to PRA
- Uncertainty ACRS meeting on Integration of Source Term Activities in Support of Advanced 28 108 Reactor Initiatives, 02/17/2022
Accident Consequence Analysis for Advanced Reactors
- Mechanistic or deterministic evaluation
- LMP assumes MST and use of PRA
- Some non-LWRs may choose to provide a postulated MHA, similar to non-power reactor licensees
- RG 1.183, regulatory position C.2, Attributes of an Acceptable AST, may be useful
- SECY-93-092 included staff recommendations on non-LWR source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced 29 109 Reactor Initiatives, 02/17/2022
Mechanistic Source Term
- SECY-93-092 definition of MST A mechanistic source term is the result of an analysis of fission product release based on the amount of cladding damage, fuel damage, and core damage resulting from the specific accident sequences being evaluated. It is developed using best-estimate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs.
ACRS meeting on Integration of Source Term Activities in Support of Advanced 30 110 Reactor Initiatives, 02/17/2022
SECY-93-092: Provisions for Staff Assurance
- The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis.
Sufficient data should exist on the reactor and fuel performance through the research, development, and testing programs to provide adequate confidence in the mechanistic approach.
- The transport of fission products can be adequately modeled for all barriers and pathways to the environs, including specific consideration of containment design. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured.
- The events considered in the analyses to develop the set of source terms for each design are selected to bound severe accidents and design-dependent uncertainties ACRS meeting on Integration of Source Term Activities in Support of Advanced 31 111 Reactor Initiatives, 02/17/2022
National Lab Non-LWR Source Term Reports
- Technology inclusive, what to do to develop accident source terms, not specific on how to do it
- No specific methods or phenomenological models
- Do not provide technology-related source terms or releases ACRS meeting on Integration of Source Term Activities in Support of Advanced 32 112 Reactor Initiatives, 02/17/2022
Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities INL/EXT-20-58717, Revision 0, June 2020, ML20192A250
- Summarizes a risk-informed, performance-based, and technology-inclusive approach to determine source terms
- Graded process
- Conservative non-mechanistic approach
- MST calculation methods
- Design-specific scenarios for a range of licensing basis events
- Best-estimate models with uncertainty quantification ACRS meeting on Integration of Source Term Activities in Support of Advanced 33 113 Reactor Initiatives, 02/17/2022
MST Formulation
=
Figure 1-2 INL/EXT-20-58717, Revision 1. From Illustration of radionuclides retention and removal process for one non-LWR concept (reproduced from SAND2020-0402)
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Technology-Inclusive Source Term Methodology Determination ACRS meeting on Integration of Source Term Activities in Support of Advanced 35 115 Reactor Initiatives, 02/17/2022
INL Report Methodology Steps 1: Identify Regulatory 8. Establish Adequacy of MST Requirements Simulation Tools 2: Identify Reference Facility 9. Develop and Update PRA Design Model 3: Define Initial Radionuclide 10. Identify or Revise the List of Inventories LBEs
- 4. Perform Bounding Calculations 11. Select LBEs to Include Design
- 5. Conduct SHA and Perform Basis External Hazard Level for Simplified Calculations Source Term Analysis
- 6. Consider Risk-informed System 12. Perform Source Term Design Changes Modeling and Simulation for LBEs
- 7. Select Initial List of LBEs and 13. Review LBEs List for Adequacy Conduct PIRT of Regulatory Acceptance
- 14. Document Completion of Source Term Development ACRS meeting on Integration of Source Term Activities in Support of Advanced 36 116 Reactor Initiatives, 02/17/2022
Simplified Approach for Scoping Assessment of Non-LWR Source Terms SAND2020-0402, January 2020, ML20052D133
- Primarily qualitative means to identify the dominant considerations that affect a release mitigation strategy
- Classifies release mitigation strategies based on a range of barriers, physical attenuation processes, and system performance under sample accident scenarios
- Did NOT develop quantitative estimates of radiological release magnitudes and compositions to the environment
- Looked at high temperature gas reactors, sodium fast reactors, and liquid fueled molten salt reactors ACRS meeting on Integration of Source Term Activities in Support of Advanced 37 117 Reactor Initiatives, 02/17/2022
Non-LWR PRA Standard ASME/ANS RA-S-1.4-2021
- Full scope PRA (includes consequence analysis)
- Mechanistic Source Term Analysis (MS) element provides useful information on what to do to develop mechanistic source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced 38 118 Reactor Initiatives, 02/17/2022
Licensing Modernization
- Risk-informed approach to selection and analysis of licensing basis events
- Combined with assessment of cumulative risks
- Key roles for PRA and MST ACRS meeting on Integration of Source Term Activities in Support of Advanced 39 119 Reactor Initiatives, 02/17/2022
Licensing Modernization See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities ACRS meeting on Integration of Source Term Activities in Support of Advanced 40 120 Reactor Initiatives, 02/17/2022
Licensing Modernization
- Flexibility provided on how to develop safety case
- NRC Advanced Reactor Policy Statement encourages use of passive and inherent features ACRS meeting on Integration of Source Term Activities in Support of Advanced 41 121 Reactor Initiatives, 02/17/2022
Assessment Frameworks Fuel Qualification (FQ)
- Top-down approach to identify criteria (goals) to support a finding that fuel is qualified ACRS meeting on Integration of Source Term Activities in Support of Advanced 42 122 Reactor Initiatives, 02/17/2022
FQ Assessment Framework Goal: Fuel is qualified for use A fuel manufacturing specification controls the key fabrication parameters that significantly affect fuel Safety criteria can be satisfied [G2]
performance [G1]
ACRS meeting on Integration of Source Term Activities in Support of Advanced 43 123 Reactor Initiatives, 02/17/2022
G2: Safety Criteria Safety criteria can be satisfied [G2]
Margin to design limits can be Margin to radionuclide demonstrated under conditions Ability to achieve and release limits under accident of normal operation, including maintain safe shutdown can conditions can be the effects of anticipated be assured [G2.3]
demonstrated [G2.2]
operational occurrences [G2.1]
10 CFR 50.34(a)(1)(ii)(D) GDC/ARDC 2 GDC/ARDC 10 10 CFR 52.47(a)(2)(iv) GDC 27/ARDC 26 10 CFR 52.79(a)(1)(vi) GDC/ARDC 35 ACRS meeting on Integration of Source Term Activities in Support of Advanced 44 124 Reactor Initiatives, 02/17/2022
G2.2: Radionuclide Release Limits Margin to radionuclide release limits under accident conditions can be demonstrated [G2.2]
Radionuclide retention The fuel performance requirements of the fuel Radionuclide retention envelope is defined under accident and release behavior of Criteria for barrier degradation the fuel matrix under
[G2.1.1] conditions is and failure under accident specified [G2.2.1] accident conditions is conditions are suitably modeled conservatively conservative [G2.2.2] [G2.2.3]
ACRS meeting on Integration of Source Term Activities in Support of Advanced 45 125 Reactor Initiatives, 02/17/2022
G2.2.2 Criteria for Barrier Degradation Criteria for barrier degradation and failure under accident conditions are suitably conservative [G2.2.2]
Criteria are shown to provide conservative Experimental data is prediction of barrier appropriate degradation and failure [G2.2.2(b)]
[G2.2.2(a)]
Note: Testing at environmental conditions consistent with accident conditions is expected (e.g., elevated temperatures)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 46 126 Reactor Initiatives, 02/17/2022
Complete FQ Assessment Framework GOAL Fuel is qualified for use GOAL Evaluation model is acceptable for use G1 Fuel is manufactured in accordance with a specification EM G1 Evaluation model contains the appropriate modeling capabilities G1.1 Key dimensions and tolerances of fuel components are specified EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system G1.2 Key constituents are specified with allowance for impurities EM G1.2 Evaluation model is capable of modeling the material properties of the fuel G1.3 End state attributes for materials within fuel components are specified or system otherwise justified EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel G2 Margin to safety limits can be demonstrated performance G2.1 Margin to design limits can be demonstrated under conditions of normal EM G2 Evaluation model has been adequately assessed against experimental data operation and AOOs EM G2.1 Data used for assessment are appropriate (see ED Assessment G2.1.1 Fuel performance envelope is defined Framework)
G2.1.2 Evaluation model is available (see EM Assessment Framework) EM G2.2 Evaluation model is demonstrably able to predict fuel failure and G2.2 Margin to radionuclide release limits under accident conditions can be degradation mechanisms over the test envelope demonstrated EM G2.2.1 Evaluation model error is quantified through assessment G2.1.1 Fuel performance envelope is defined against experimental data G2.2.1 Radionuclide retention requirements are specified EM G2.2.2 Evaluation model error is determined throughout the fuel G2.2.2 Criteria for barrier degradation and failure are suitably conservative performance envelope EM G2.2.3 Sparse data regions are justified (a) Criteria are conservative EM G2.2.4 Evaluation model is restricted to use within its test envelope (b) Experimental data are appropriate (see ED Assessment Framework)
G2.2.3 Radionuclide retention and release from fuel matrix are modeled GOAL Experimental data used for assessment are appropriate conservatively ED G1 Assessment data are independent of data used to develop/train the evaluation model (a) Model is conservative ED G2 Data has been collected over a test envelope that covers the fuel performance (b) Experimental data are appropriate (see ED Assessment envelope Framework) ED G3 Experimental data have been accurately measured G2.3 Ability to achieve and maintain safe shutdown is assured ED G3.1 The test facility has an appropriate quality assurance program G2.3.1 Coolable geometry is ensured ED G3.2 Experimental data are collected using established measurement techniques (a) Criteria to ensure coolable geometry are specified ED G3.3 Experimental data account for sources of experimental uncertainty (b) Evaluation models are available (see EM Assessment ED G4 Test specimens are representative of the fuel design Framework) ED G4.1 Test specimens are fabricated consistent with the fuel manufacturing G2.3.2 Negative reactivity insertion can be demonstrated specification (a) Criteria are provided to ensure that negative reactivity ED G4.2 Distortions are justified and accounted for in the experimental data insertion path is not obstructed (b) Evaluation model is available (see EM Assessment Framework)
- For illustrative purposes only. Please see Appendix A to NUREG-2246 for a legible list.
ACRS meeting on Integration of Source Term Activities in Support of Advanced 47 127 Reactor Initiatives, 02/17/2022
Non-LWR Accident Source Term Webpage Information https://www.nrc.gov/reactors/new-reactors/advanced/related-documents/nuclear-power-reactor-source-term.html
- One-stop shop for existing information, on public website under advanced reactors
- Discussion of accident source terms
- Linked list of documents relevant to development of non-LWR accident source terms for licensing
- Staff will keep up to date ACRS meeting on Integration of Source Term Activities in Support of Advanced 48 128 Reactor Initiatives, 02/17/2022
Acronyms AST alternative source term EPZ emergency planning zone INL Idaho National Laboratory LBE licensing basis event LMP Licensing Modernization Project LWR light water reactor MHA maximum hypothetical accident MST mechanistic source term Non-LWR non-light water reactor PIRT phenomena identification and ranking table PRA probabilistic risk assessment RG regulatory guide SHA system hazard analysis ACRS meeting on Integration of Source Term Activities in Support of Advanced 49 129 Reactor Initiatives, 02/17/2022
Guidance for developing advanced reactor source term (long-term)
Bill Reckley Michelle Hart John Segala NRR/DANU 130
General Approach
- Technology-inclusive methodology available as an option
- Actual implementation is technology/design specific
- NRC not planning to provide analytical inputs to applicants (beyond making available NRC developed models)
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DOE/National Laboratories ACRS meeting on Integration of Source Term Activities in Support of Advanced 3 132 Reactor Initiatives, 02/17/2022
NRC Activities ACRS meeting on Integration of Source Term Activities in Support of Advanced 4 133 Reactor Initiatives, 02/17/2022
Next Generation Nuclear Plant (NGNP)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 5 134 Reactor Initiatives, 02/17/2022
Model Development ACRS meeting on Integration of Source Term Activities in Support of Advanced 6 135 Reactor Initiatives, 02/17/2022
Applications & Pre-App Interactions ACRS meeting on Integration of Source Term Activities in Support of Advanced 7 136 Reactor Initiatives, 02/17/2022
Moving Forward
- Following the scientific work being done by national laboratories and developers
- Engaging with developers
- Continuing to develop NRC models and identify related uncertainties
- Consider additional guidance based on experience with ongoing interactions
- Consider feedback on the new webpage ACRS meeting on Integration of Source Term Activities in Support of Advanced 8 137 Reactor Initiatives, 02/17/2022
Opportunity for Public Comment ACRS meeting on Integration of Source Term Activities in Support of Advanced 138 Reactor Initiatives, 02/17/2022
Member Discussion ACRS meeting on Integration of Source Term Activities in Support of Advanced 139 Reactor Initiatives, 02/17/2022
Adjourn ACRS meeting on Integration of Source Term Activities in Support of Advanced 140 Reactor Initiatives, 02/17/2022
Full Name User Action Timestamp Snodderly, Michael Joined 2/17/2022, 9:00:15 AM Walton, Shandeth Joined before 2/17/2022, 9:00:15 AM Burkhart, Larry Joined before 2/17/2022, 9:00:15 AM Dashiell, Thomas Joined before 2/17/2022, 9:00:15 AM Dave Petti (Guest) Joined before 2/17/2022, 9:00:15 AM Costa, Arlon Joined 2/17/2022, 9:04:17 AM Nourbakhsh, Hossein Joined 2/17/2022, 9:05:11 AM Smith, Micheal Joined 2/17/2022, 9:07:45 AM Smith, Micheal Left 2/17/2022, 9:07:55 AM Smith, Micheal Joined 2/17/2022, 9:25:52 AM Dennis Bley (Guest) Joined 2/17/2022, 9:09:23 AM Widmayer, Derek Joined 2/17/2022, 9:10:42 AM James Cordes (Guest) Joined 2/17/2022, 9:12:14 AM Kyriazidis, Lucas Joined 2/17/2022, 9:27:04 AM Blumberg, Mark Joined 2/17/2022, 9:13:03 AM Sunseri, Matthew Joined 2/17/2022, 9:15:00 AM Segala, John Joined 2/17/2022, 9:18:06 AM Hart, Michelle Joined 2/17/2022, 9:18:57 AM Dickson, Elijah Joined 2/17/2022, 9:19:59 AM Halnon, Gregory Joined 2/17/2022, 9:20:29 AM Jose March-Leuba (ACRS) (GuJoined 2/17/2022, 9:20:52 AM Rempe, Joy Joined 2/17/2022, 9:21:23 AM Grady, Anne-Marie Joined 2/17/2022, 9:21:25 AM Schaperow, Jason Joined 2/17/2022, 9:23:04 AM Reckley, William Joined 2/17/2022, 9:23:26 AM Bergman, Jana Joined 2/17/2022, 9:24:59 AM Ronald G Ballinger Joined 2/17/2022, 9:25:03 AM Wang, Weidong Joined 2/17/2022, 9:25:38 AM Bier, Vicki Joined 2/17/2022, 9:26:21 AM Meighan, Sean Joined 2/17/2022, 9:26:32 AM Kirchner, Walter Joined 2/17/2022, 9:27:03 AM Marksberry, Don Joined 2/17/2022, 9:27:26 AM Vesna B. Dimitrijevic Joined 2/17/2022, 9:28:05 AM Schultz, Stephen Joined 2/17/2022, 9:28:10 AM Brown, Charles Joined 2/17/2022, 9:28:24 AM Hsueh, Kevin Joined 2/17/2022, 9:28:31 AM Drew Peebles (Kairos Power) Joined 2/17/2022, 9:28:35 AM Parillo, John Joined 2/17/2022, 9:29:02 AM Rautzen, William Joined 2/17/2022, 9:29:03 AM AUSTGEN, Kati Joined 2/17/2022, 9:29:10 AM AUSTGEN, Kati Left 2/17/2022, 10:01:13 AM Clement, Richard Joined 2/17/2022, 9:29:25 AM Sarah Fields (Guest) Joined 2/17/2022, 9:29:48 AM Helvenston, Edward Joined 2/17/2022, 9:29:52 AM Drzewiecki, Timothy Joined 2/17/2022, 9:29:54 AM Campbell, Shawn Joined 2/17/2022, 9:29:58 AM Kock, Andrea Joined 2/17/2022, 9:29:58 AM Gran, Zachary Joined 2/17/2022, 9:30:02 AM
Tharakan, Binesh Joined 2/17/2022, 9:30:11 AM Salay, Michael Joined 2/17/2022, 9:30:15 AM Esmaili, Hossein Joined 2/17/2022, 9:30:18 AM Nakanishi, Tony Joined 2/17/2022, 9:30:51 AM Nakanishi, Tony Left 2/17/2022, 10:15:50 AM Philpott, Stephen Joined 2/17/2022, 9:31:26 AM Ghosh, Tina Joined 2/17/2022, 9:31:35 AM Tetter, Keith Joined 2/17/2022, 9:32:08 AM Stutzcage, Edward Joined 2/17/2022, 9:32:09 AM Franovich, Mike Joined 2/17/2022, 9:32:15 AM Gus Merwin (Kairos Power) (G Joined 2/17/2022, 9:32:27 AM Bielen, Andrew Joined 2/17/2022, 9:32:49 AM Hayes, Michelle Joined 2/17/2022, 9:34:17 AM Luxat, David Lyle Joined 2/17/2022, 9:34:34 AM Sebrosky, Joseph Joined 2/17/2022, 9:35:30 AM Lynch, Steven Joined 2/17/2022, 9:35:44 AM Oesterle, Eric Joined 2/17/2022, 9:38:39 AM Palmrose, Donald Joined 2/17/2022, 9:39:20 AM Palmrose, Donald Left 2/17/2022, 9:50:39 AM Valliere, Nanette Joined 2/17/2022, 9:39:32 AM Wagner, Kenneth Charles Joined 2/17/2022, 9:40:26 AM Jung, Ian Joined 2/17/2022, 9:41:46 AM Garry, Steven Joined 2/17/2022, 9:44:02 AM Webber, Kimberly Joined 2/17/2022, 9:44:32 AM Garmon-Candelaria, David Joined 2/17/2022, 9:44:42 AM Cubbage, Amy Joined 2/17/2022, 9:44:46 AM Vasavada, Shilp Joined 2/17/2022, 9:46:29 AM Pohida, Marie Joined 2/17/2022, 9:50:24 AM Wieselquist, William Joined 2/17/2022, 9:51:36 AM Hoellman, Jordan Joined 2/17/2022, 9:53:58 AM Walker, Shakur Joined 2/17/2022, 9:55:00 AM Christopher P. Chwasz Joined 2/17/2022, 9:57:54 AM Bostelmann, Rike Joined 2/17/2022, 10:09:29 AM Donoghue, Joseph Joined 2/17/2022, 10:09:52 AM Bryan Eyers (TerraPower) (Gu Joined 2/17/2022, 10:16:39 AM