ML21356A607

Transcript for the Advisory Committee on Reactor Safeguards Kairos Power Licensing Subcommittee Meeting - November 19, 2021, Pages 1-61 (Open)

ML21356A607
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Issue date:	11/19/2021
From:	Advisory Committee on Reactor Safeguards
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Wang, W, ACRS
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NRC-1754
Download: ML21356A607 (61)

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Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION

Title:

Advisory Committee on Reactor Safeguards Kairos Power Licensing Subcommittee Open Session Docket Number: (n/a)

Location: teleconference Date: Friday, November 19, 2021 Work Order No.: NRC-1754 Pages 1-41 NEAL R. GROSS AND CO., INC.

Court Reporters and Transcribers 1716 14th Street, N.W., Suite 200 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 KAIROS POWER LICENSING SUBCOMMITTEE 8 + + + + +

9 FRIDAY 10 NOVEMBER 19, 2021 11 + + + + +

12 The Subcommittee met via Teleconference, 13 at 9:00 a.m. EST, David A. Petti, Chair, presiding.

14 15 COMMITTEE MEMBERS:

16 DAVID A. PETTI, Chair 17 RONALD G. BALLINGER, Member 18 VICKI M. BIER, Member 19 DENNIS BLEY, Member 20 CHARLES H. BROWN, JR., Member 21 VESNA B. DIMITRIJEVIC, Member 22 JOSE MARCH-LEUBA, Member 23 JOY L. REMPE, Member 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

2 1 ACRS CONSULTANT:

2 STEPHEN SCHULTZ 3

4 DESIGNATED FEDERAL OFFICIAL:

5 WEIDONG WANG 6

7 8

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3 1 T-A-B-L-E O-F C-O-N-T-E-N-T-S 2 ACRS Chairman Introductory Remarks . . . . . . . 4 3 NRC Staff Introductory Remarks . . . . . . . . . 6 4 Kairos Power Introductory Remarks . . . . . . . . 10 5 Overview of Kairos Power Mechanistic Source Term 6 (MST) Methodology . . . . . . . . . . . . . . . 13 7 Public Comment . . . . . . . . . . . . . . . . . 41 8 Adjournment . . . . . . . . . . . . . . . . . . . 41 9

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4 1 P-R-O-C-E-E-D-I-N-G-S 2 9:00 a.m.

3 CHAIR PETTI: Well, good morning, 4 everyone. The meeting will now come to order.

5 This is a meeting of the Kairos Power 6 Licensing Subcommittee of the Advisory Committee on 7 Reactor Safeguards. I am David Petti, Chairman of 8 today's Subcommittee meeting.

9 ACRS members in attendance are Vicki Bier, 10 Charles Brown, Jose March-Leuba, Joy Rempe, Ron 11 Ballinger, Vesna Dimitrijevic. And I don't see 12 anybody else.

13 Our consultants. Let's see. I don't see 14 any of our consultants either at this point, but they 15 may be in by phone.

16 Weidong Wang of the ACRS staff is the 17 designated federal official for this meeting.

18 During today's meeting the Subcommittee 19 will review staff Safety Evaluation Report on the KP-20 FHR Mechanistic Source Term Methodology, Revision 1.

21 The Subcommittee will hear presentations by and hold 22 discussions with the NRC staff, Kairos Power 23 representatives, and other interested persons 24 regarding this matter, but part of the presentations 25 by the applicant and the NRC staff may be closed 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

5 1 order to discuss information that is proprietary to 2 the licensee and its contractors pursuant to 5 U.S.C.

3 552(b)(C)(4).

4 Attendance at the meeting that deals with 5 such information will be limited to the NRC staff and 6 its consultants, Kairos Power, and those individuals 7 and organizations who have entered into an appropriate 8 confidentiality agreement with them. Consequently we 9 will need to confirm that we have only eligible 10 observers and participants in the closed part of the 11 meeting.

12 The rules for participation in all ACRS 13 meetings including today's were announced in the 14 Federal Register on June 13th, 2019. The ACRS section 15 of the U.S. NRC public website provides our charter, 16 bylaws, agendas, letter reports, and full transcripts 17 of all Full and Subcommittee meetings including slides 18 presented there. The meeting notice and agenda for 19 this meeting were posted there. We have received no 20 written statements or requests to make an oral 21 statement from the public.

22 The Subcommittee will gather information, 23 analyze relevant issues and facts, and formulate 24 proposed positions and actions as appropriate for 25 deliberation by the Full Committee.

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6 1 The rules for participation in today's 2 meetings have been announced as part of the notice of 3 this meeting previously published in the Federal 4 Register.

5 A transcript of the meeting is being kept 6 and will be made available as stated in the Federal 7 Register notice.

8 Due to the COVID pandemic today's meeting 9 is being held over Microsoft Teams for ACRS, NRC 10 staff, and the licensee attendees. There is also a 11 telephone bridge line allowing participation of the 12 public over the phone.

13 When addressing the Subcommittee that 14 participant should first identify themselves and speak 15 with sufficient clarity and volume so that they may be 16 readily heard. When not speaking we request that 17 participants mute your computer microphone or phone.

18 We'll now proceed with the meeting. And 19 I'd like to start by calling on William Kennedy, NRR 20 management.

21 MR. KENNEDY: Well, good morning, Mr.

22 Chairman and distinguished members of the Advisory 23 Committee on Reactor Safeguards. My name is William 24 Kennedy. I'm the Acting Chief of the Advanced Reactor 25 Licensing Branch in NRR's Division of Advanced 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 reactors and Non-Power Production and Utilization 2 Facilities.

3 It's my pleasure to be here today to 4 provide introductory remarks on behalf of the 5 division. With me today are Ms. Michelle Hart, who is 6 the lead technical reviewer. Mr. Alex Chereskin.

7 They are both from the Advanced Reactor Technical 8 Branch No. 2 in DANU. Mr. Jason White is here from 9 the External Hazards Branch in the Division of 10 Engineering and External Hazards. And all of them 11 will be providing the staff presentation. We also 12 have Mr. Samuel Cuadrado de Jesus who is providing 13 project management support for the review of this 14 topical report.

15 The staff is looking forward to 16 discussions with and feedback from ACRS members today 17 on the Draft Safety Evaluation of the Kairos Power 18 topical report that's titled KP-FHR Mechanistic Source 19 Term Methodology. So as you will hear this topical 20 report is important for Kairos' development of 21 accident source terms and atmospheric dispersion 22 values for use in radiological consequence analysis 23 for siting and safety analysis for Kairos Power's 24 fluoride salt-cooled high-temperature reactor designs, 25 also known as the KP-FHR designs.

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8 1 The report also describes development of 2 source terms for estimation of dose for anticipated 3 operational occurrences in design-basis events to be 4 used in the endorsed NEI 18-04 methodology for 5 applicants to categorize events, classify and describe 6 special treatment of structures, systems, and 7 components, and assess defense-in-depth for non-light 8 water reactors.

9 This topical report is related to other 10 Kairos topical reports such as the Fuel Performance 11 Methodology Report. Limitations and conditions on the 12 use of the topical report are identified to ensure 13 that the methods and underlying assumptions are 14 applicable to the specific design in future KP-FHR 15 license applications.

16 So I'd just like to note that this is the 17 fourth time the staff and Kairos Power have had the 18 opportunity to brief ACRS on Kairos' topical reports 19 and so the staff appreciated the helpful comments from 20 the ACRS on the recent topical report evaluation 21 covering reactor coolant scaling methodologies, 22 licensing modernization project implementation, and 23 most recently the Fuel Performance Methodology Report.

24 Staff looks forward to continuing to work 25 with Chairman Petti and the rest of the ACRS members 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 and staff as we complete reviews of more Kairos Power 2 topical reports and review license applications for 3 facilities that will use the Kairos Power design.

4 In September we received a construction 5 permit application for the Kairos Power Hermes test 6 reactor, and that is currently being reviewed for 7 acceptance.

8 I'd also like to highlight the working 9 relationship between the NRC staff and Kairos Power 10 has been excellent. Similar to previous reviews of 11 Kairos Power topical reports the staff and Kairos have 12 used public meetings as an efficient means for 13 addressing technical issues without the need for 14 significant formal requests for additional 15 information.

16 And then finally I'd like to give a big 17 thanks to the technical staff for their efforts to 18 produce a high-quality Draft Safety Evaluation Report 19 and also for project management of this review.

20 So that concludes my opening remarks.

21 Thank you very much.

22 CHAIR PETTI: Thank you. Before we turn 23 it over to Kairos I just want to note for the record 24 that our consultant Steve Schultz has joined us.

25 So, Kairos, the floor is yours.

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10 1 MEMBER BROWN: Dave, are they showing 2 their slides?

3 CHAIR PETTI: Not yet.

4 Kairos, are you out there?

5 MEMBER BROWN: Just wanted to make sure I 6 wasn't the only one.

7 MR. PEEBLES: Okay. Sorry. We were 8 having some technical difficulties with the conference 9 room.

10 Thank you, Mr. Chairman, and good morning, 11 everyone. My name is Drew Peebles. I'm the Manager 12 of Licensing and Safety Integration here at Kairos 13 Power. Before we get started I would like to thank 14 the ACRS members (audio interference).

15 CHAIR PETTI: Okay. I don't hear them 16 anymore. Do other people have that problem?

17 MEMBER BROWN: It sounds like we've lost 18 them, Dave.

19 MEMBER MARCH-LEUBA: Yes, like -- yes, 20 they were having -- we were trying to test their 21 conference room yesterday. They were having some 22 technical issues.

23 CHAIR PETTI: Okay.

24 (Pause.)

25 CHAIR PETTI: Okay. I hear you guys 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

11 1 again.

2 MR. PEEBLES: Sorry about that. So I was 3 saying we would like to thank the ACRS members for 4 your continued interest in Kairos Power. William 5 Kennedy mentioned that we've had four briefings in 6 front of the ACRS to date. I believe this is the 7 fifth topical that we will bring to you on.

8 (Audio interference.)

9 CHAIR PETTI: And they've faded out again.

10 So some of the -- I'm assuming some of the 11 Kairos folks that I see listed that may not be in the 12 conference room are texting them and telling them.

13 (Pause.)

14 MR. PEEBLES: Sorry about this. So I 15 think I mentioned that William Kennedy also mentioned 16 that we were -- that we've briefed the ACRS four times 17 to date and I believe this is the fifth topical that 18 we get a chance (audio interference).

19 CHAIR PETTI: Okay. We're continuing to 20 have problems. I'm wondering if I should go out and 21 come back in, if that would help.

22 MR. PEEBLES: Okay. Can you hear us now?

23 CHAIR PETTI: A little echo, but yes. Oh, 24 a big echo.

25 MR. PEEBLES: Okay. We've joined with 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

12 1 different laptop. Sorry about the conference room 2 issues.

3 So as I was mentioning, we have recently 4 submitted our construction permit application for our 5 non-power reactor that we refer to as Hermes and we 6 look forward to engaging with the ACRS in the review 7 of that application as well.

8 I would also like to thank the NRC staff 9 for a thorough and efficient review of the topical 10 report. I think all of the feedback and discussions 11 made sure that we had a complete product.

12 So I'm joined here by the lead technical 13 contributor to the topical report, Dr. Matthew Denman, 14 who will be giving the presentation today. We are 15 also joined by several subject matter experts that 16 will be available to answer detailed questions in 17 their areas of expertise.

18 Just as a reminder to the Kairos 19 attendees, if you do come off mute, please remember to 20 introduce yourselves.

21 And with that I will turn it over to Matt.

22 And let me make sure I can share my slides on this 23 computer.

24 DR. DENMAN: Yes. So while Drew is 25 pulling the slides up I'll begin my introductions.

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13 1 Mr. Chairman, members of the Committee, 2 thank you very much for your time today. My name is 3 Matthew Denman and I am a principal reliability 4 engineer at Kairos Power and it is going to be my 5 pleasure today to brief you on Kairos Power's 6 Mechanistic Source Term Methodology Topical Report.

7 And, Weidong, can you make sure that I'm 8 a presenter so I can share my screen?

9 MR. WANG: I think you are. You are the 10 presenter.

11 DR. DENMAN: The --

12 MR. WANG: Maybe you -- a different --

13 okay. Now it's because --

14 DR. DENMAN: Yes.

15 MR. WANG: -- you changed it up. Okay.

16 Let me just go and make -- yes, it's changed.

17 (Pause.)

18 MR. WANG: We can see your screen now.

19 DR. DENMAN: Thank you so much.

20 Okay. So with that Kairos Powers' mission 21 is to enable the world's transition to clean energy 22 with the ultimate goal of dramatically improving 23 people's quality of life while protecting the 24 environment.

25 In order to achieve this mission we must 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 prioritize that our efforts focus on a clean energy 2 technology that is both affordable and safe. Today's 3 topic, mechanistic source term, is key to allowing 4 Kairos Power to demonstrate the safety of our design 5 which will enable the affordability of that design.

6 At a high level our approach to source 7 term is to decompose the problem into a series of 8 material-at-risk throughout the plant and barrier 9 release fractions that will separate that material at 10 risk from our receptor at the site boundary.

11 For each barrier radionuclides are grouped 12 and then we model the release of that group of 13 radionuclides through the barrier using a 14 representative element. Barriers for radionuclide 15 release are the TRISO fuel and the FLiBe coolant.

16 These form our functional containment for 17 radionuclides.

18 Again radionuclide groupings are used to 19 facilitate transport of radionuclides through the 20 barriers and unique grouping structures will exist for 21 various release models. So for say the fuel you might 22 have a different grouping structure for mechanical 23 grinding of the fuel verse a diffusion of 24 radionuclides through the TRISO barriers.

25 At steady --

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15 1 CHAIR PETTI: Matthew?

2 DR. DENMAN: Yes, sir.

3 CHAIR PETTI: I just have a real high-4 level question here on the methodology.

5 DR. DENMAN: Sure.

6 CHAIR PETTI: I understand it's to be used 7 really for accidents, but do you guys plan to use this 8 same methodology to support the worker dose 9 evaluations, shielding needs, or are you guys thinking 10 about a completely different approach there?

11 DR. DENMAN: That is a very good question 12 and thank you very much for it. The approach in the 13 topical is limited to off-site dose calculations and 14 explicitly excludes worker does or control room dose.

15 Similar methods may be used to quantify those dose 16 metrics, but the complete strategy of how to drive a 17 conservative consequence estimate has not been 18 included in this topical report.

19 CHAIR PETTI: Okay. Thanks.

20 DR. DENMAN: So for sources of steady 21 state material at risk in the system the overwhelming 22 majority of our material at risk is contained within 23 our TRISO fuel. That TRISO fuel can exist in multiple 24 configurations. Most of our TRISO fuel will exist as 25 either completely intact particles or with 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

16 1 compromised inner or outer PyC layer and all of these 2 configurations are expected to retain an overwhelming 3 majority of the fission products and heavy metals 4 contained within those particles.

5 There will be -- due to manufacturing and 6 in-service, steady state in-service failures there 7 will be some TRISO particles that will have 8 compromised silicon carbide layers, and these 9 particles are expected to release a certain quantity 10 of their fission products into the FLiBe coolant 11 during steady state irradiation.

12 Additionally, as part of the manufacturing 13 process there is a very small fraction of dispersed 14 uranium that is expected throughout the fuel form and 15 the fission products from this dispersed uranium have 16 no credited fission product or heavy metal retention 17 capabilities within the fuel.

18 These fission products and heavy metals 19 will move into the circulating activity where they 20 will be combined with impurities that are expected 21 within the salt including sodium, uranium, thorium, 22 various other corrosion products.

23 The circulating activity will continue to 24 generate radionuclides via transmutation. There will 25 be some tritium production within the FLiBe coolant.

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17 1 The tritium primarily will either be absorbed into 2 graphite pebbles and structural materials or will move 3 into various off-gas cleanup systems.

4 When we talk about how we're going to 5 quantify the material at risk throughout the plant, 6 for the fuel we will focus on our manufacturing 7 specifications. We will utilize the KP-BISON Fuel 8 Performance Code to estimate the depletion of 9 radionuclides from the fuel and we will use our Core 10 Design Topical Report methodology in order to 11 calculate the burnup and buildup of fission products 12 within the fuel.

13 The circulating activity material at risk 14 will be limited by technical specifications that will 15 be set as limiting conditions of operations for our 16 plant.

17 The holdup of tritium in structures and 18 graphite pebbles will be calculated via the tritium 19 source term methodology discussed in the next few 20 slides. And various material at risks outside of our 21 functional containment will be limited by its 22 technical specifications, specifically the FLiBe 23 cleanup and -- one sec.

24 (Pause.)

25 DR. DENMAN: Sorry. My apologies 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

18 1 those technical difficulties.

2 The material at risk within the various 3 cleanup systems will be limited by technical 4 specifications.

5 For steady state tritium inventory 6 evaluations, tritium will be -- or tritium modeling 7 will include transport and holdup in the fuel pebbles 8 and core moderator and graphite structures in the 9 vessel and primary piping and intermediate heat 10 exchange steel. Tritium is produced in the KP-FHR 11 through the reactions listed below. The top two 12 reactions are the primary reactions that contribute to 13 tritium production in the system and the bottom two 14 reactions are the primary reactions contributing to 15 lithium-6 buildup, which will subsequently be sources 16 of tritium production.

17 CHAIR PETTI: So, Matt, just another 18 question. So you're not explicitly modeling tritium 19 production in the graphite from lithium impurities nor 20 ternary fission in the particles?

21 DR. DENMAN: So, well, we are --

22 CHAIR PETTI: I mean they may be 23 significantly smaller here but --

24 DR. DENMAN: Yes.

25 CHAIR PETTI: -- it's probably worth just 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 a -- I think they'll be smaller given the capability 2 to make good graphite today. Years ago impurities 3 were higher and you had to worry about those things.

4 DR. DENMAN: Understand. Yes, we are not 5 explicitly modeling the lithium and the graphite, nor 6 the ternary fission within the fuel due to the fact 7 that an overwhelming majority of the tritium that is 8 expected to be produced in the system will be produced 9 via the FLiBe reactions shown on the slide.

10 CHAIR PETTI: Okay. Thanks.

11 DR. DENMAN: The KP-FHR is uniquely suited 12 to retain radionuclides due to the large margins to 13 fuel damage from our operating range. Our core inlet 14 and outlet temperatures are in the 550 to 650 range.

15 Our FLiBe freezing temperatures and our -- sorry, our 16 FLiBe boiling temperatures are not until 1,430 degrees 17 C. And our peak fuel temperatures above which we 18 would potentially expect silicon carbide-induced 19 failures aren't until 1,600 degrees C. So there is a 20 large margin to the functional failure of our various 21 radionuclide barriers within our functional 22 containment approach.

23 For MAR mobilization for anticipated 24 operational occurrences, design-basis events and 25 design-basis accidents it should be emphasized 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

20 1 under design -- or under the conditions expected for 2 these events only a minor fraction of the total 3 material at risk in our plant can potentially be 4 mobilized because a majority of our material at risk 5 is contained safely within our TRISO fuel.

6 We expect no incremental fuel failures 7 below 1,600 degrees C and there are multiple inherent 8 safety features in our design to protect the fuel from 9 achieving such high temperatures.

10 The material at risk in the reactor 11 coolant as well as the material at risk presented in 12 other locations can be mobilizing in anticipated 13 operational occurrences, design-basis events, and 14 design-basis accidents particularly via aerosolization 15 of the FLiBe such as for jet breakup in a hypothetical 16 guillotine rupture of a primary pipe or vaporization 17 of chemical species within the FLiBe at elevated 18 temperatures, although there are only limited release 19 rates expected due to evaporation of soluble 20 radionuclides from FLiBe at temperatures below 816 21 degrees C, which is our vessel limit and sets the 22 upper bound of our design-basis --

23 CHAIR PETTI: Matt?

24 DR. DENMAN: Yes, sir.

25 CHAIR PETTI: Just the first sub-bullet 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

21 1 the first bullet, the way it's stated. The EPRI 2 topical report has a failure fraction at -- what's 3 called at high temperature. It's a statistical zero 4 level because the testing showed there were no 5 failures. Are you assuming that level for any 6 accident event or are you saying zero is zero?

7 DR. DENMAN: We will use the KP-BISON Fuel 8 Performance Topical Report to calculate the stresses 9 and strains on the various barriers and the 10 incremental fuel failure fraction. It is our 11 expectation that that value will be near zero, below 12 1,600 degrees C, but our methodology is to actually 13 calculate that.

14 CHAIR PETTI: So I have the same problem 15 with the last topical. If that number is lower than 16 what has been measured statistically, how do you 17 validate that number?

18 DR. DENMAN: I will pass this question 19 along to our fuel performance expert Ryan Latta.

20 Ryan, can you jump on?

21 MR. LATTA: Hello?

22 DR. DENMAN: Yes, Ryan?

23 MR. LATTA: Yes, this is Ryan Latta. Yes, 24 the current methodology is to use the Fuel Performance 25 Code. And it calculates the radiation history, uses 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 the radiation history's input, and then goes through 2 the transient analysis and uses the actual conditions 3 that are for the accident, which are significantly 4 below conditions that were tested for the furnace 5 annealing test. So we probably have a 4 to 500 degree 6 margin from the conditions that were in the furnace 7 safety testing. So when you follow that track you end 8 up with very low, near negligible failure fractions 9 during an accident event. And so that's how the --

10 that's the methodology we followed for --

11 DR. DENMAN: And I will add -- this is 12 Matthew Denman again. I will add that the methodology 13 for determining where the -- or which barriers are 14 intact are failed lies squarely within the fuel 15 performance methodology. This topical report on 16 source term basically looks only at -- once you've 17 determined which barriers are available for release, 18 how do you move radionuclides through those barriers?

19 So we kind of take the configuration of 20 the TRISO fuel as a given boundary condition from the 21 Fuel Performance Topical Report.

22 CHAIR PETTI: All right.

23 MEMBER REMPE: This is Joy. And first of 24 all, I'd like to ask people who aren't speaking to 25 mute their computers or phones because there's a lot 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

23 1 of background noise I'm hearing, but a couple of 2 questions.

3 I know the topical report says you don't 4 -- now I'm getting an echo. So again, people, please 5 mute. Okay?

6 But anyway, the topical report says you're 7 not going to deal with beyond-design-basis events, but 8 yet several times it talks about well, you'll just 9 continue things for beyond-design-basis events. So 10 could you clarify, are you planning to go ahead and 11 use these same models and extend them for beyond-12 design-basis events or are you going to use a 13 different methodology?

14 And then I didn't ask earlier, but I was 15 curious, the topical report continues to say that, as 16 other ones did, the coolant is an important barrier 17 for release. And it doesn't talk about the fact that 18 the coolant can interact with other barriers and 19 degrade them. And how are you planning to modify this 20 methodology to consider this degradation?

21 I'm sorry. Did -- I'm not hearing any 22 response, so maybe now it's time to un-mute, whoever 23 is trying to talk or respond.

24 MEMBER BLEY: Joy, I can hear you, so they 25 ought to.

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24 1 MEMBER REMPE: Thank you for that 2 confirmation, but I was asking a lot of people to mute 3 so maybe they haven't un-muted yet.

4 DR. DENMAN: Yes, I think I got muted 5 without my knowledge. My apologies.

6 Thank you very much, Joy, for those 7 questions. I will answer the beyond-design-basis 8 question first.

9 So the methodologies that we developed for 10 this topical report were the methodologies from 11 phenomena that we expected to exist in anticipated 12 operational occurrences, design-basis events, and 13 design-basis accident boundary conditions. It is 14 possible that in beyond-design-basis space that you 15 will experience similar boundary conditions, and in 16 those cases the methodologies may be extended into 17 beyond-design-basis conditions. However, there are 18 expected to be additional scenarios in beyond-design-19 basis event space that extend beyond the applicability 20 of these models, and at that point we would have to 21 revise and justify the models in a future license 22 application. Does that answer your --

23 MEMBER REMPE: So you are planning to 24 extend or do something with KP-BISON? You're not just 25 going to say okay, now we're going to go and use 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 something else that's similar to MELCOR or something 2 like that, or you've just not decided yet what tool 3 you'll use?

4 MR. PEEBLES: So this is Drew Peebles with 5 Licensing. So not in this topical. So we would 6 definitely deal with that in the future application of 7 the methodology. So if we do extend beyond the 8 design-basis, then we would have to justify how we're 9 doing that in that future license application. But 10 for this particular topical report we weren't asking 11 for an NRC finding on beyond-design-basis conditions.

12 MEMBER REMPE: Again, I understand that 13 you've said that you're limiting to design-basis 14 events, but then in the report it continues to make 15 reference to beyond-design-basis events and I'm not 16 getting an answer to the question are you going to use 17 this tool or another tool, or you've not decided what 18 tool --

19 DR. DENMAN: Yes, I think the short answer 20 is we haven't decided upon the beyond-design-basis --

21 (Simultaneous speaking.)

22 MEMBER REMPE: Okay. And then what about 23 degradation due to long-term operation, from corrosion 24 or something between the coolant, which you continue 25 to say you think is a barrier, but one thing 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

26 1 unusual about this design is that it's -- there's the 2 potential that some of the barriers can degrade other 3 barriers. And how are you planning to modify -- I 4 didn't see anything discussed about how you will 5 simulate that phenomena in this topical report.

6 DR. DENMAN: Thank you very much for that 7 question, Joy. Particularly for the fuel in transient 8 conditions we do not expect under very short time 9 horizons for there to be induced failure of the fuel 10 barrier such that you would have fuel/FLiBe 11 interactions, and that is explicitly called out in the 12 topical report.

13 Under longer term conditions if there were 14 to be fuel/FLiBe interactions, then the radionuclides 15 from the fuel would move into the FLiBe and join the 16 circulating activity. And we have a technical 17 specification on circulating activity, so as long as 18 the circulating activity remains below that technical 19 specification, our methodology would still hold.

20 MEMBER REMPE: So --

21 DR. DENMAN: Matthew. My apologies.

22 MEMBER REMPE: So you're basically saying 23 you don't model degradation with long-term operation.

24 You just think it's not going to be that important as 25 long as the coolant circulating activity stays below 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 a certain value?

2 DR. DENMAN: Correct.

3 MEMBER REMPE: That you're just not 4 simulating that phenomena? Then what about can the 5 circulating activity, if it were to start degrading 6 other subsequent barriers due to corrosion of some of 7 the structural material? Are you still -- are you 8 also going to be neglecting it? And then you'll -- is 9 this something that's built into the model, you 10 constantly do a check to make sure the circulating 11 activity stays below that value all the time so that 12 you don't ever exceed this? So is that something 13 you've put into KP-BISON to do some sort of check?

14 DR. DENMAN: So the circulating activity 15 technical specification will be a limiting condition 16 of operation. We will monitor the circulating 17 activity over the life time of the reactor operations 18 and ensure that we are below the value set forth in 19 our license.

20 MEMBER REMPE: We're talking about the 21 tool today. And so you're telling me well okay, so 22 the tool doesn't have to consider this degradation 23 interaction between the coolant barrier and the 24 barriers within the fuel. So basically you're saying 25 that if you're doing a simulation to provide some sort 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

28 1 of source term to the NRC, you're constantly doing 2 some sort of check to make sure that you don't -- if 3 you're going to do source term after long-term 4 operation, end of cycle, that you've done a check 5 always in the tool to make sure it's below that value, 6 right? Is what you're telling me?

7 DR. DENMAN: Thank you very much. Not 8 quite. We will set a technical specification in our 9 license application that sets the upper limit of 10 circulating activity in our FLiBe. We will use KP-11 BISON to model normal buildup and diffusion of 12 radionuclides out of the fuel, but that only sets the 13 initial condition of material at risk within the fuel 14 itself.

15 In the FLiBe for any accident condition we 16 will use the technical specification -- or any design-17 basis accident condition we will use the technical 18 specification value as the initial condition of 19 circulating activity in the FLiBe. So we will not be 20 calculating in an a priori estimating what that 21 release would be. We will use the upper bound value 22 of acceptable circulating activity as our initial 23 condition for the accident.

24 MEMBER REMPE: Okay. And then what about 25 if there's interactions between the FLiBe 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

29 1 structural material?

2 DR. DENMAN: So the Structures Topical 3 Report will ensure that the vessel is not degraded 4 with -- by the FLiBe. Within our anticipated 5 operational range any other structure system and 6 component is not safety-related, and breaks in those 7 systems would be evaluated in our postulated event 8 analysis.

9 MEMBER REMPE: Okay. Thank you.

10 DR. DENMAN: So as a part of this analysis 11 we're not explicitly modeling that.

12 MEMBER REMPE: Okay. Thank you.

13 CHAIR PETTI: Just a clarification. So 14 the tech spec on circulating activity, is that 15 basically equivalent to what the gas reactor guys are 16 talking about SARDL?

17 DR. DENMAN: They're related concepts, 18 although we do not believe that we would set a limit 19 on the circulating activity that would be the break 20 point between acceptable or unacceptable off-site 21 doses. We would choose a value that is -- that we 22 believe is achievable to be monitored and measurable 23 and ensure the safety of the system. But it might be 24 slightly -- formulated in a slightly different way 25 that the SARDLs.

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

30 1 CHAIR PETTI: Okay.

2 DR. DENMAN: Okay. So and then there's 3 also going to be tritium that's going to be stored in 4 the graphite pebbles and structures that can be 5 desorbed at elevated temperatures and our methodology 6 will examine that release.

7 Our design-basis accident site boundary 8 dose will be used -- dose is going to demonstrate that 9 the KP-FHR meets dose limits in 10 C.F.R. 50.34, 10 52.79, and 100.11. Again, technical specifications 11 will be set on the activity of the FLiBe, cover gas 12 and other systems, and the system will be design to 13 preclude incremental fuel failures from DBA conditions 14 as evaluated by KP-BISON.

15 Anticipated operational occurrences and 16 design-basis event source term analyses are similar to 17 design-basis accidents, but more realistic assessments 18 of barriers, mitigation strategies and initial 19 conditions may be assumed.

20 The circulating activity technical 21 specification will be used to inform operational 22 limits on circulating activity. These operational 23 limits may be more realistic conditions for normal 24 operation effluent calculations as well as anticipated 25 operational occurrences and 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

31 1 Our radionuclide grouping and transport 2 approach is very similar to that used in light water 3 reactor safety analysis. I have the MELCOR grouping 4 structure on the right here where you can see the 5 various chemical groups and then the representative 6 element at the top that represents now the releases 7 from those groups are calculated for light water 8 reactor. We take a similar approach, although we 9 evaluate the grouping structures specifically to the 10 barrier and the release mode within that barrier.

11 So essentially our approach is we look at 12 individual isotopes within a barrier and combine them 13 into their RN groups, their radionuclide groups. We 14 calculate the release fractions for each radionuclide 15 group associated with the medium as calculated by 16 driving forces within that barrier: temperatures, 17 pressures. Release fractions are combined with 18 relevant inventories to determine the quantity of that 19 material that is mobilized.

20 Once you move from one barrier to the next 21 the radionuclide inventory is combined with any 22 radionuclides that are already present in that next 23 barrier and then regrouped for subsequent 24 mobilization. Once you reach the gas space the dose 25 consequences for radionuclides that are transferred 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 into the gas space are evaluated with RADTRAD and 2 ARCON.

3 CHAIR PETTI: So, Matthew, I had bad 4 network quality there for a minute so I missed it, but 5 the groupings are the same no matter where the fission 6 product is in the system, or does it -- when it's in 7 the fuel it's considered one way because of the 8 chemistry there. When it's in the salt it's 9 considered because of the chemistry there?

10 DR. DENMAN: Correct. Every barrier will 11 have its unique grouping structure. And specifically 12 for the fuel there is a unique grouping structure for 13 diffusion versus mechanical grinding of the fuel. So 14 different release pathways may have their own unique 15 grouping structure compared to the -- and then each 16 barrier will have its own unique grouping structure.

17 CHAIR PETTI: Okay.

18 DR. DENMAN: Okay. Our primary barrier 19 for radionuclide retention is our TRISO fuel. This 20 fuel contains an overwhelming majority of the material 21 at risk within our plant during normal and off-normal 22 operating modes. Again, a series of diverse and 23 robust barriers to radionuclide retention with 24 extensive industrial fabrication experience and 25 irradiation under a variety of conditions such as 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

33 1 Advanced Gas Reactor Development Program as mentioned 2 earlier.

3 Our TRISO fuel manufacturing 4 specifications will determine how the fuel 5 configurations begin in the transient. Fission 6 products will diffuse from imperfect particles, 7 primarily particles with failed silicon carbon layers 8 or that have exposed kernels. Radionuclides from 9 heavy metal contamination from the manufacturing 10 process will have no credit for radionuclide retention 11 in steady state.

12 Minimum expected steady state diffusion of 13 radionuclides are expected from the remaining 14 configurations with intact silicon carbide layers.

15 Compromised configurations will partially or entirely 16 be depleted fission products during steady state thus 17 reducing the available material at risk within those 18 TRISO configurations during the transient.

19 For the FLiBe barrier this is the second 20 part of our functional containment for radionuclide 21 retention. Once radionuclides are in the FLiBe they 22 will be separated into either salt soluble compounds, 23 suspended oxides, noble metals or gases. And Kairos 24 Power's Fuel Development Program, or sorry, FLiBe 25 Development Program builds on radionuclide retention 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 experience in the molten salt reactor experiment with 2 the exception that our salt is going to be much, much 3 cleaner than what was experienced in MSRE, which was 4 a fuel salt system.

5 CHAIR PETTI: So, Matthew, just a question 6 on that, and if I get into proprietary stuff, just 7 tell me and we'll cover it in the closed session.

8 I noticed that you had put some fissile 9 impurities in the salt, and I was surprised at that 10 level being that high. And I wasn't sure if that was 11 just being conservative or what was done back in the 12 old days of MSRE or whether that actually is what you 13 get.

14 My experience in gas reactors is in the 15 old days stuff was just not as clean as you can get 16 today with today's technology and I wasn't sure 17 whether this was a holdover from that. I would have 18 thought you'd probably be able to get better, cleaner 19 salt than that.

20 DR. DENMAN: So the cleanest of the salt 21 is going to be dependent upon the economics of the 22 system and how much we want to pay for various grades.

23 Those decisions are not made at this point in time and 24 our methodology is designed to be flexible enough to 25 account for various levels of impurities.

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 CHAIR PETTI: So that's sort of a -- let's 2 just say a conservative level. That may not be what 3 you actually see in practice.

4 DR. DENMAN: Correct.

5 CHAIR PETTI: Okay.

6 DR. DENMAN: For tritium transport, 7 tritium transport within structures is determined by 8 mass transfer coefficients from FLiBe flow 9 characteristics throughout the system. Transport 10 within structures is determined based upon material 11 properties such as diffusion within and through steel, 12 diffusion and trapping within pebbles and structural 13 graphite.

14 Salt structure boundary conditions set by 15 material tritium -- is set by material tritium 16 solubility, particularly Henry's Law for solubility of 17 tritium fluoride and tritium gas in FLiBe, and 18 Sievert's Law for solubility of tritium in steel.

19 CHAIR PETTI: So, Matthew, just a comment 20 here. The amount of literature on tritium behavior in 21 these materials, both the salt and the graphitic 22 material, is quite large and there's a lot of 23 uncertainty. These measurements are not easy to make.

24 Diffusion in liquids are notoriously difficult and 25 have high uncertainty. Solubilities are not easy.

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36 1 It is now understood better that simple 2 experiments where one injects tritium into molten 3 coolants like this show a bias. The tritium doesn't 4 actually go in. It can sit along the surface. If you 5 think of like a loop. This has been shown in Europe 6 in the Fusion Program for a different coolant that's 7 a low-solubility coolant. FLiBe is a low-solubility 8 coolant.

9 And I think it's very difficult. This is 10 exactly how I would model it. I just think the 11 validation is going to be quite challenging because 12 the experiments may have these biases that you --

13 until you get to the actual in situ generation of 14 tritium, you may be surprised. And it's just 15 something that I think when you're doing sensitivity 16 studies on the model you got to open up the window 17 here because there's a lot of stuff that even though 18 the experimentalists have done the best that they can 19 do, without in situ generating tritium it's really 20 difficult.

21 In terms of the graphitic material I would 22 again caution that the fusion experiences on graphites 23 that are not these graphites. Pebble graphitic 24 material is not a graphite, whereas the -- your 25 reflective material is a nuclear graphite. Those 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

37 1 very different microstructures so that there's very 2 big differences potentially in trapped concentrations.

3 I believe the trapped energies are probably generic to 4 carbon materials, but the actual concentrations are 5 very strongly microstructural-dependent. Radiation 6 can affect it, too. All of these things make it much 7 more complicated than these really nice elegant 8 models.

9 And if you go back -- you have to go back 10 a little ways in the fusion world to see some of the 11 models and the differences and some of the complexity 12 there. It's just a caution that when you think about 13 the validation, you think about sensitivity, keep the 14 window open large because of these differences.

15 I also recommend that if you haven't 16 looked at complexity of models, take a look at the 17 modeling that's done to date. There has been recent 18 publications on air ingress with graphite. Oak Ridge 19 and Idaho have done a tremendous amount of modeling, 20 highly complex, and they try to bring in the 21 microstructure. And it takes you back to say, wow, 22 there's a lot there. They spent a decade getting all 23 the parameters that you need to really understand it.

24 And then you look at these models which 25 are much simpler and it just gives you a cause 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

38 1 concern. So it's worth looking at some of that as you 2 think about how you're going to bound things and do 3 sensitivity analysis and the like.

4 DR. DENMAN: Thank you very much for your 5 feedback on that. It's very valuable and insightful 6 and we'll take it as we move forward with this 7 approach.

8 For tritium from the FLiBe-free surface, 9 tritium fluoride and tritium gas can both exist as 10 dissolved gases in the FLiBe. Contributions to off-11 site dose would either require permeation into vessel 12 or piping and then release into the reactor building 13 or evolution into the gas space which is modeled via 14 the gas transport equations influenced by the 15 experiments as shown below.

16 For gas space analysis we are using the 17 NRC codes RADTRAD and ARCON96. These are used to 18 model radionuclides traveling through the building and 19 off site. And to support dose calculations the 20 existing models and framework set forth in these codes 21 are accepted as is.

22 For RADTRAD as input we need the mobilized 23 material at risk from the previous barriers as 24 previously discussed. RADTRAD will handle all the 25 radionuclide decay and for the entire duration 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

39 1 transient and use the Henry correlation for aerosol 2 settling, and conservatively and prescriptively 3 leakage rates will be applied out of the reactor 4 building into the environment.

5 For ARCON the release definitions around 6 the release of radioactive material from the site the 7 location of the receptor and meteorological conditions 8 at the site are needed to calculate chi over qs.

9 Various limitations are set forth in this 10 topical report. They are listed on the slide, but I 11 will not read them word for word.

12 And with that are there any further 13 questions?

14 CHAIR PETTI: Just another comment in the 15 tritium realm with the nitrate salt. Whenever one is 16 dealing with lower levels of tritium, there's always 17 a waste management concern. You get to a point where 18 the concentrations are so low it's hard to find a 19 disposal route. The folks in EDA (phonetic) have been 20 struggling with this. When you have lots of tritium, 21 there's lots of technologies to be able to concentrate 22 it, move it, get it where you want it to be, put it on 23 a bed or something, but when you get to low 24 concentrations, it's above what's allowed to be 25 released, but it's so low that the technologies 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

40 1 deal with it are a problem. So it's just something to 2 put on your tickler list as your design evolves.

3 Again my information may be a little out 4 of date, but this was at least the case ten years ago, 5 but they were still struggling with some of these 6 sorts of issues.

7 DR. DENMAN: Thank you very much for that 8 feedback. It's definitely something that we'll take 9 back as we continue to mature our design.

10 I'm not able to see the chat window or 11 anything, so if there's any further questions?

12 CHAIR PETTI: Yes, members, any questions?

13 DR. DENMAN: Well, hearing none, I really 14 appreciate your time in this open session and look 15 forward to continued conversations in the closed 16 session.

17 CHAIR PETTI: Okay. Thanks.

18 Is Michelle going to talk? Who's going to 19 talk for the staff?

20 MR. CUADRADO DE JESUS: For the staff we 21 don't have presentations for the open session.

22 CHAIR PETTI: Ah, okay. Then I guess with 23 that we can move to the closed session.

24 MR. WANG: Dave?

25 CHAIR PETTI: Yes.

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41 1 MR. WANG: We need to have public comment.

2 CHAIR PETTI: Yes, yes, yes. So okay, 3 let's open -- anybody that has a comment from the 4 public, *6 to un-mute yourself. Give us your name and 5 your comment.

6 Okay. Hearing none, I guess we will end 7 this open session. And I think all the members should 8 have the link to the closed session.

9 And, Kairos, we'll want you to make sure 10 that all the folks you think should be there should be 11 there and Weidong and our staff will handle the NRC 12 side.

13 So with that we'll see everybody in the 14 closed session.

15 (Whereupon, the above-entitled matter went 16 off the record at 9:56 a.m.)

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

KP-NRC-2111-001 Enclosure 2 Open Session Presentation Slides for the November 19, 2021 ACRS Kairos Power Subcommittee Briefing (Non-Proprietary)

KP-FHR Mechanistic Source Term Methodology Topical Report ACRS Subcommittee Meeting, November 19, 2021 Copyright © 2021 Kairos Power LLC. All Rights Reserved.

No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

Kairos Powers mission is to enable the worlds transition to clean energy, with the ultimate goal of dramatically improving peoples quality of life while protecting the environment.

In order to achieve this mission, we must prioritize our efforts to focus on a clean energy technology that is affordable and safe.

Copyright © 2021 Kairos Power LLC. All Rights Reserved.

High Level Approach Source Term Methodology

• Decompose the problem into a series of Material at Risk (MAR) and barrier Release Fractions (RFs) that separate that MAR from a receptor at the site boundary.

• For each barrier, group radionuclides into and model release through that barrier using a representative element for that group.

The barriers for radionuclide release are the TRISO fuel and the Flibe coolant (i.e., functional containment).

Radionuclide groups are used to facilitate transport through barriers.

Unique grouping structures exist for specific release modes (e.g., mechanical grinding of fuel in the PHSS vs diffusion through TRISO barriers).

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3

Sources of Steady State Material at Risk (MAR)

Fuel Circulating Activity Flibe Cleanup Intact Layers Uranium, Noble Metals Compromised IPyC Layer Thorium, Metals Transmutation and Fission Offgas Compromised OPyC Layer Initial Salt Assumed to retain FPs and HM Loading Gases and Vapors Compromised SiC Layer Tritium Building Exposed Kernel Production FP and HMC Inservice Failures Tritium from Fuel Assumed to release FPs Structures (Graphite + Pebbles) Nitrate Dispersed Uranium No Retention of FPs or HM Tritium Tritium C.A. Contamination Copyright © 2021 Kairos Power LLC. All Rights Reserved.

No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

4

Sources of Steady State Material at Risk (MAR)

Fuel Circulating Activity Flibe Cleanup Intact Layers Uranium, Noble Technical Metals Specifications Manufacturing Specifications Compromised IPyC Layer Thorium, Metals Transmutation

+ OPyC Layer and Fission Offgas Compromised Initial Salt Assumed to retain FPs and HM Technical Specifications Loading Technical Specifications Gases and Vapors KP-BISON (FuelCompromised Performance SiCTopical)

Layer Tritium Building Exposed Kernel Production

+ FP and HMC Inservice Failures TechnicalTritium Specifications from Fuel Assumed to release FPs Burnup (Core Design Topical) Structures (Graphite + Pebbles) Nitrate Dispersed Uranium No Retention of FPs or HM Tritium TechnicalTritium Specifications Tritium C.A. Contamination Source Term Methodology Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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5

Steady State Tritium Inventory

• Tritium modeling will include transport and holdup in:

Fuel Pebbles and core moderator Graphite Structures Vessel Steel Primary piping Intermediate Heat Exchangers

• Tritium is produced in the KP-FHR through the following reactions:

Modeled Tritium Production Reactions Modeled Li-6 build-in Reactions Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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6

KP-FHR Specifications Uniquely Large Margins Between Operational and Failure Temperatures Parameter Value/Description Reactor Type Fluoride-salt cooled, high temperature reactor (FHR)

Core Configuration Pebble bed core, graphite moderator/reflector, and enriched Flibe molten salt coolant Core Inlet and Exit Temperature 550°C / 600-650°C Design Temperature Limits Value Primary Salt (Flibe) Freezing and Boiling Temperatures 459°C / 1430°C Maximum ASME Section III, Division 5, SS316 Temperature 816°C Peak Fuel Temperature Limit 1600°C Our combination of fuel and coolant provides a uniquely large safety margin.

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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

7

MAR Mobilization in AOOs, DBEs, and DBAs Only minor fractions of the total MAR can be mobilized in AOOs, DBEs, or DBAs

• The vast majority of MAR is safely protected in the fuel during AOOs, DBEs, and DBAs.

No incremental fuel failure is expected at temperatures <1600C.

Multiple inherent safety features protect the fuel from achieving high temperatures.

• MAR circulating in the reactor coolant as well as MAR present in other locations (cover gas, intermediate loop, etc) can be mobilized in AOOs, DBEs, and DBAs.

Aerosolization of Flibe - Hypothetical guillotine pipe break or primary pump operations Vaporization- chemical specific evaporation is evaluated across accident temperature profiles Limited release rates are expected from evaporation of soluble radionuclides from Flibe for temperatures below 816C.

• Tritium stored in graphite, pebbles, and structures can be desorbed at elevated temperatures.

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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

8

AOO, DBE, & DBA Source Term Methodology

• DBA site boundary dose to demonstrate KP-FHR meets dose limits in 10 CFR 50.34, 10 CFR 52.79, and 10 CFR 100.11.

• A technical specification (tech spec) limit will be set on activity in the Flibe, cover gas, and other systems.

The system is designed to preclude incremental fuel failures due to the DBA conditions as evaluated by KP-BISON.

• AOO and DBE source term analyses similar to DBAs, but a more realistic assessment of barriers, mitigation strategies, and initial conditions may be assumed.

• The circulating activity technical spec. will be used to inform an operational limit on circulating activity. This operational limit can be used as a more realistic initial condition for normal operation effluent calculations as well as certain AOOs and DBEs.

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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

9

Radionuclide Grouping and Transport Approach

• Transport of radionuclides through each LWR Example medium is evaluated on an RN group basis using the following steps:

1. Individual isotopes are combined into RN group for each barrier.

2. Release fractions of each RN group associated with that medium is calculated given driving forces (e.g., temperature, pressure).

3. Release fractions are combined with the relevant inventories to determine the quantity of material that is mobilized. That incoming material is then:

1. Combined with the radionuclides already present in the next barrier and then

2. Regrouped for subsequent mobilization

4. The dose consequences for radionuclides that are transferred into the gas space are evaluated with RADTRAD and ARCON.

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10

KP-FHR Fuel Element The Primary Barrier of Radionuclide Retention

• The TRISO fuel form provides the first barrier to radionuclide retention in the KP-FHR for all normal and off-normal operating modes

• TRISO particles utilize a series of diverse barriers to provide robust fuel performance

• Kairos Powers fuel design builds on the AGR fuel development program Extensive industrial fabrication experience Validated irradiation performance under a wide variety of conditions Copyright © 2021 Kairos Power LLC. All Rights Reserved.

No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

11

TRISO Fuel Configurations for Material at Risk (MAR)

• Fuel manufacturing specifications will determine what radionuclides Fuel begin the transient in the fuel:

Intact Layers Fission products (FPs) diffusion from imperfect particles Compromised IPyC Layer Compromised SiC layer but intact IPyC and/or OPyC will release a smaller a range of FPs to the Flibe. Compromised OPyC Layer Exposed kernels and in-service TRISO failures mobilize mobile a larger range FPs Assumed to retain FPs and HM transported to the Flibe at steady state.

Radionuclides from heavy metal contamination from the manufacturing process will Compromised SiC Layer have no credited retention at steady state. Exposed Kernel Minimal expected steady state diffusion of radionuclides are expected from Inservice Failures the remaining configurations.

Assumed to release FPs

• The compromised configurations will be partially or entirely depleted Dispersed Uranium of FPs during steady state, thus reducing the MAR available for release during the transient. No Retention of FPs or HM TRISO Cohorts Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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12

KP-FHR Flibe Coolant The Second Barrier of Radionuclide Retention

• The primary coolant, Flibe, provides a secondary functional containment barrier to radionuclide retention in the KP-FHR for all in-core normal and off-normal operating modes.

• Flibe can chemically react with fission and activation products, separating them into:

salt soluble compounds suspended oxides noble metals, or Impurities gas phases

• Kairos Powers Flibe development program builds on radionuclide retention experience in the Molten Salt Reactor Experiment Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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13

Tritium Uptake into Structures Metallics, Fuel and Moderator Pebbles, Structural Graphite

• Tritium transport to structures determined by mass transfer coefficients from Flibe flow characteristics

• Transport within structures determined based on material properties Tritium diffusion modeled within steels Tritium diffusion + trapping modeled within pebbles and structural graphite

• Salt/Structure boundary condition set by material tritium solubility Henrys law solubility for TF and T2 in Flibe Sieverts law solubility for T in steel Examples:

Heat Exchanger - Permeation through Metal Core - Retention in Pebbles Downcomer - Vessel Permeation and Reflector Retention Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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14

Tritium from Flibe Free Surface Evolution into the Cover Gas

• Tritium fluoride (TF) and Tritium (T2) can both exist as dissolved gases in Flibe.

• Contribution to offsite dose would require either:

Permeation into vessel or piping and then releasing to the reactor building or Evolution into the cover gas which is modeled using mass transport equations influenced by experiments conducted by Suzuki et al.

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Gas Space Analysis Simple and Conservative

• Codes: RADTRAD and ARCON96 Gas space transport KP MST Models Dose calculations Existing models are accepted as-is.

• Key Inputs:

RADTRAD: Conservative and Prescriptive Values ARCON Mobilized material-at-risk activities Depletion mechanisms Radioactive decay and/or Henry correlation for aerosol settling.

Leakage rates (Conservative)

ARCON:

Release definitions Receptor definitions Meteorological data Copyright © 2021 Kairos Power LLC. All Rights Reserved.

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Limitations

• 1. Approval of KPBison for use in fuel performance analysis as captured in KP-TR-010-P (KP-FHR Fuel Performance Methodology).

• 2. Justification of thermodynamic data and associated vapor pressure correlations of representative species.

• 3. Validation of tritium transport modeling methodology.

• 4. Confirmation of minimal ingress of Flibe into pebble matrix carbon under normal and accident conditions, such that incremental damage to TRISO particles due to chemical interaction does not occur as captured in KP-TR-010-P (Fuel Qualification Methodology for the KP-FHR).

• 5. Establishment of operating limitations on maximum circulating activity and concentrations relative to solubility limits in the reactor coolant, intermediate coolant, cover gas, and radwaste systems that are consistent with the initial condition assumptions in the safety analysis report.

• 6. Quantification of the transport of tritium in nitrate salt and between nitrate salt and the cover gas

• 7. The phenomena associated with radionuclide retention discussed in this report is restricted to molten Flibe.

The retention of radionuclides in solid Flibe is beyond the scope of the current analysis.

• 8. The methodology presented in this report is based on design features of a KPFHR (details provided in report).

Deviations from these design features will be justified by an applicant in safety analysis reports associated with license application submittals.

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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.

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