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Transcript for the Advisory Committee on Reactor Safeguards Kairos Power Licensing Subcommittee Meeting - November 19, 2021, Pages 1-61 (Open)
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Issue date:	11/19/2021
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	Wang, W, ACRS
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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.

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1 2

3 DISCLAIMER 4

5 6

UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8

9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.

15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.

19 20 21 22 23

1 UNITED STATES OF AMERICA 1

NUCLEAR REGULATORY COMMISSION 2

+ + + + +

3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4

(ACRS) 5

+ + + + +

6 KAIROS POWER LICENSING SUBCOMMITTEE 7

+ + + + +

8 FRIDAY 9

NOVEMBER 19, 2021 10

+ + + + +

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

13 14 COMMITTEE MEMBERS:

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

2 ACRS CONSULTANT:

1 STEPHEN SCHULTZ 2

3 DESIGNATED FEDERAL OFFICIAL:

4 WEIDONG WANG 5

6 7

8 9

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3 T-A-B-L-E O-F C-O-N-T-E-N-T-S 1

ACRS Chairman Introductory Remarks 4

2 NRC Staff Introductory Remarks 6

3 Kairos Power Introductory Remarks........ 10 4

Overview of Kairos Power Mechanistic Source Term 5

(MST) Methodology............... 13 6

Public Comment

................. 41 7

Adjournment................... 41 8

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4 P-R-O-C-E-E-D-I-N-G-S 1

9:00 a.m.

2 CHAIR PETTI:

Well, good

morning, 3

everyone. The meeting will now come to order.

4 This is a meeting of the Kairos Power 5

Licensing Subcommittee of the Advisory Committee on 6

Reactor Safeguards. I am David Petti, Chairman of 7

today's Subcommittee meeting.

8 ACRS members in attendance are Vicki Bier, 9

Charles Brown, Jose March-Leuba, Joy Rempe, Ron 10 Ballinger, Vesna Dimitrijevic. And I don't see 11 anybody else.

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

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

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

20 The Subcommittee will hear presentations by and hold 21 discussions with the NRC

staff, Kairos Power 22 representatives, and other interested persons 23 regarding this matter, but part of the presentations 24 by the applicant and the NRC staff may be closed in 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

5 order to discuss information that is proprietary to 1

the licensee and its contractors pursuant to 5 U.S.C.

2 552(b)(C)(4).

3 Attendance at the meeting that deals with 4

such information will be limited to the NRC staff and 5

its consultants, Kairos Power, and those individuals 6

and organizations who have entered into an appropriate 7

confidentiality agreement with them. Consequently we 8

will need to confirm that we have only eligible 9

observers and participants in the closed part of the 10 meeting.

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

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

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

6 The rules for participation in today's 1

meetings have been announced as part of the notice of 2

this meeting previously published in the Federal 3

Register.

4 A transcript of the meeting is being kept 5

and will be made available as stated in the Federal 6

Register notice.

7 Due to the COVID pandemic today's meeting 8

is being held over Microsoft Teams for ACRS, NRC 9

staff, and the licensee attendees. There is also a 10 telephone bridge line allowing participation of the 11 public over the phone.

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

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

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

21 Chairman and distinguished members of the Advisory 22 Committee on Reactor Safeguards. My name is William 23 Kennedy. I'm the Acting Chief of the Advanced Reactor 24 Licensing Branch in NRR's Division of Advanced 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

7 reactors and Non-Power Production and Utilization 1

Facilities.

2 It's my pleasure to be here today to 3

provide introductory remarks on behalf of the 4

division. With me today are Ms. Michelle Hart, who is 5

the lead technical reviewer. Mr. Alex Chereskin.

6 They are both from the Advanced Reactor Technical 7

Branch No. 2 in DANU. Mr. Jason White is here from 8

the External Hazards Branch in the Division of 9

Engineering and External Hazards. And all of them 10 will be providing the staff presentation. We also 11 have Mr. Samuel Cuadrado de Jesus who is providing 12 project management support for the review of this 13 topical report.

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

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8 The report also describes development of 1

source terms for estimation of dose for anticipated 2

operational occurrences in design-basis events to be 3

used in the endorsed NEI 18-04 methodology for 4

applicants to categorize events, classify and describe 5

special treatment of structures,

systems, and 6

components, and assess defense-in-depth for non-light 7

water reactors.

8 This topical report is related to other 9

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

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

23 Staff looks forward to continuing to work 24 with Chairman Petti and the rest of the ACRS members 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

9 and staff as we complete reviews of more Kairos Power 1

topical reports and review license applications for 2

facilities that will use the Kairos Power design.

3 In September we received a construction 4

permit application for the Kairos Power Hermes test 5

reactor, and that is currently being reviewed for 6

acceptance.

7 I'd also like to highlight the working 8

relationship between the NRC staff and Kairos Power 9

has been excellent. Similar to previous reviews of 10 Kairos Power topical reports the staff and Kairos have 11 used public meetings as an efficient means for 12 addressing technical issues without the need for 13 significant formal requests for additional 14 information.

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

19 So that concludes my opening remarks.

20 Thank you very much.

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

24 So, Kairos, the floor is yours.

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10 MEMBER BROWN: Dave, are they showing 1

their slides?

2 CHAIR PETTI: Not yet.

3 Kairos, are you out there?

4 MEMBER BROWN: Just wanted to make sure I 5

wasn't the only one.

6 MR. PEEBLES: Okay. Sorry. We were 7

having some technical difficulties with the conference 8

room.

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

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

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

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

22 CHAIR PETTI: Okay.

23 (Pause.)

24 CHAIR PETTI: Okay. I hear you guys 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

11 again.

1 MR. PEEBLES: Sorry about that. So I was 2

saying we would like to thank the ACRS members for 3

your continued interest in Kairos Power. William 4

Kennedy mentioned that we've had four briefings in 5

front of the ACRS to date. I believe this is the 6

fifth topical that we will bring to you on.

7 (Audio interference.)

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

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

12 (Pause.)

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

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

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

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

24 MR. PEEBLES: Okay. We've joined with a 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

12 different laptop. Sorry about the conference room 1

issues.

2 So as I was mentioning, we have recently 3

submitted our construction permit application for our 4

non-power reactor that we refer to as Hermes and we 5

look forward to engaging with the ACRS in the review 6

of that application as well.

7 I would also like to thank the NRC staff 8

for a thorough and efficient review of the topical 9

report. I think all of the feedback and discussions 10 made sure that we had a complete product.

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

17 Just as a

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

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

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

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

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13 Mr. Chairman, members of the Committee, 1

thank you very much for your time today. My name is 2

Matthew Denman and I am a principal reliability 3

engineer at Kairos Power and it is going to be my 4

pleasure today to brief you on Kairos Power's 5

Mechanistic Source Term Methodology Topical Report.

6 And, Weidong, can you make sure that I'm 7

a presenter so I can share my screen?

8 MR. WANG: I think you are. You are the 9

presenter.

10 DR. DENMAN: The --

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

12 okay. Now it's because --

13 DR. DENMAN: Yes.

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

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

16 (Pause.)

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

18 DR. DENMAN: Thank you so much.

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

24 In order to achieve this mission we must 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

14 prioritize that our efforts focus on a clean energy 1

technology that is both affordable and safe. Today's 2

topic, mechanistic source term, is key to allowing 3

Kairos Power to demonstrate the safety of our design 4

which will enable the affordability of that design.

5 At a high level our approach to source 6

term is to decompose the problem into a series of 7

material-at-risk throughout the plant and barrier 8

release fractions that will separate that material at 9

risk from our receptor at the site boundary.

10 For each barrier radionuclides are grouped 11 and then we model the release of that group of 12 radionuclides through the barrier using a

13 representative element. Barriers for radionuclide 14 release are the TRISO fuel and the FLiBe coolant.

15 These form our functional containment for 16 radionuclides.

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

diffusion of 23 radionuclides through the TRISO barriers.

24 At steady --

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15 CHAIR PETTI: Matthew?

1 DR. DENMAN: Yes, sir.

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

4 DR. DENMAN: Sure.

5 CHAIR PETTI: I understand it's to be used 6

really for accidents, but do you guys plan to use this 7

same methodology to support the worker dose 8

evaluations, shielding needs, or are you guys thinking 9

about a completely different approach there?

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

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

18 CHAIR PETTI: Okay. Thanks.

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

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16 compromised inner or outer PyC layer and all of these 1

configurations are expected to retain an overwhelming 2

majority of the fission products and heavy metals 3

contained within those particles.

4 There will be -- due to manufacturing and 5

in-service, steady state in-service failures there 6

will be some TRISO particles that will have 7

compromised silicon carbide

layers, and these 8

particles are expected to release a certain quantity 9

of their fission products into the FLiBe coolant 10 during steady state irradiation.

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

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

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

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17 The tritium primarily will either be absorbed into 1

graphite pebbles and structural materials or will move 2

into various off-gas cleanup systems.

3 When we talk about how we're going to 4

quantify the material at risk throughout the plant, 5

for the fuel we will focus on our manufacturing 6

specifications. We will utilize the KP-BISON Fuel 7

Performance Code to estimate the depletion of 8

radionuclides from the fuel and we will use our Core 9

Design Topical Report methodology in order to 10 calculate the burnup and buildup of fission products 11 within the fuel.

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

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

23 (Pause.)

24 DR. DENMAN: Sorry. My apologies for 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

18 those technical difficulties.

1 The material at risk within the various 2

cleanup systems will be limited by technical 3

specifications.

4 For steady state tritium inventory 5

evaluations, tritium will be -- or tritium modeling 6

will include transport and holdup in the fuel pebbles 7

and core moderator and graphite structures in the 8

vessel and primary piping and intermediate heat 9

exchange steel. Tritium is produced in the KP-FHR 10 through the reactions listed below. The top two 11 reactions are the primary reactions that contribute to 12 tritium production in the system and the bottom two 13 reactions are the primary reactions contributing to 14 lithium-6 buildup, which will subsequently be sources 15 of tritium production.

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

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

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

23 DR. DENMAN: Yes.

24 CHAIR PETTI: -- it's probably worth just 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

19 a -- I think they'll be smaller given the capability 1

to make good graphite today. Years ago impurities 2

were higher and you had to worry about those things.

3 DR. DENMAN: Understand. Yes, we are not 4

explicitly modeling the lithium and the graphite, nor 5

the ternary fission within the fuel due to the fact 6

that an overwhelming majority of the tritium that is 7

expected to be produced in the system will be produced 8

via the FLiBe reactions shown on the slide.

9 CHAIR PETTI: Okay. Thanks.

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

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

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

20 under design -- or under the conditions expected for 1

these events only a minor fraction of the total 2

material at risk in our plant can potentially be 3

mobilized because a majority of our material at risk 4

is contained safely within our TRISO fuel.

5 We expect no incremental fuel failures 6

below 1,600 degrees C and there are multiple inherent 7

safety features in our design to protect the fuel from 8

achieving such high temperatures.

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

22 CHAIR PETTI: Matt?

23 DR. DENMAN: Yes, sir.

24 CHAIR PETTI: Just the first sub-bullet in 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

21 the first bullet, the way it's stated. The EPRI 1

topical report has a failure fraction at -- what's 2

called at high temperature. It's a statistical zero 3

level because the testing showed there were no 4

failures. Are you assuming that level for any 5

accident event or are you saying zero is zero?

6 DR. DENMAN: We will use the KP-BISON Fuel 7

Performance Topical Report to calculate the stresses 8

and strains on the various barriers and the 9

incremental fuel failure fraction. It is our 10 expectation that that value will be near zero, below 11 1,600 degrees C, but our methodology is to actually 12 calculate that.

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

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

19 Ryan, can you jump on?

20 MR. LATTA: Hello?

21 DR. DENMAN: Yes, Ryan?

22 MR. LATTA: Yes, this is Ryan Latta. Yes, 23 the current methodology is to use the Fuel Performance 24 Code. And it calculates the radiation history, uses 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

22 the radiation history's input, and then goes through 1

the transient analysis and uses the actual conditions 2

that are for the accident, which are significantly 3

below conditions that were tested for the furnace 4

annealing test. So we probably have a 4 to 500 degree 5

margin from the conditions that were in the furnace 6

safety testing. So when you follow that track you end 7

up with very low, near negligible failure fractions 8

during an accident event. And so that's how the --

9 that's the methodology we followed for --

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

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

21 CHAIR PETTI: All right.

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

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

questions.

2 I know the topical report says you don't 3

-- now I'm getting an echo. So again, people, please 4

mute. Okay?

5 But anyway, the topical report says you're 6

not going to deal with beyond-design-basis events, but 7

yet several times it talks about well, you'll just 8

continue things for beyond-design-basis events. So 9

could you clarify, are you planning to go ahead and 10 use these same models and extend them for beyond-11 design-basis events or are you going to use a 12 different methodology?

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

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

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

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24 MEMBER REMPE: Thank you for that 1

confirmation, but I was asking a lot of people to mute 2

so maybe they haven't un-muted yet.

3 DR. DENMAN: Yes, I think I got muted 4

without my knowledge. My apologies.

5 Thank you very much, Joy, for those 6

questions. I will answer the beyond-design-basis 7

question first.

8 So the methodologies that we developed for 9

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

22 MEMBER REMPE: So you are planning to 23 extend or do something with KP-BISON? You're not just 24 going to say okay, now we're going to go and use 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

25 something else that's similar to MELCOR or something 1

like that, or you've just not decided yet what tool 2

you'll use?

3 MR. PEEBLES: So this is Drew Peebles with 4

Licensing. So not in this topical. So we would 5

definitely deal with that in the future application of 6

the methodology. So if we do extend beyond the 7

design-basis, then we would have to justify how we're 8

doing that in that future license application. But 9

for this particular topical report we weren't asking 10 for an NRC finding on beyond-design-basis conditions.

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

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

20 (Simultaneous speaking.)

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

26 unusual about this design is that it's -- there's the 1

potential that some of the barriers can degrade other 2

barriers. And how are you planning to modify -- I 3

didn't see anything discussed about how you will 4

simulate that phenomena in this topical report.

5 DR. DENMAN: Thank you very much for that 6

question, Joy. Particularly for the fuel in transient 7

conditions we do not expect under very short time 8

horizons for there to be induced failure of the fuel 9

barrier such that you would have fuel/FLiBe 10 interactions, and that is explicitly called out in the 11 topical report.

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

19 MEMBER REMPE: So --

20 DR. DENMAN: Matthew. My apologies.

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

23 You just think it's not going to be that important as 24 long as the coolant circulating activity stays below 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

27 a certain value?

1 DR. DENMAN: Correct.

2 MEMBER REMPE: That you're just not 3

simulating that phenomena? Then what about can the 4

circulating activity, if it were to start degrading 5

other subsequent barriers due to corrosion of some of 6

the structural material? Are you still -- are you 7

also going to be neglecting it? And then you'll -- is 8

this something that's built into the model, you 9

constantly do a check to make sure the circulating 10 activity stays below that value all the time so that 11 you don't ever exceed this? So is that something 12 you've put into KP-BISON to do some sort of check?

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

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

28 of source term to the NRC, you're constantly doing 1

some sort of check to make sure that you don't -- if 2

you're going to do source term after long-term 3

operation, end of cycle, that you've done a check 4

always in the tool to make sure it's below that value, 5

right? Is what you're telling me?

6 DR. DENMAN: Thank you very much. Not 7

quite. We will set a technical specification in our 8

license application that sets the upper limit of 9

circulating activity in our FLiBe. We will use KP-10 BISON to model normal buildup and diffusion of 11 radionuclides out of the fuel, but that only sets the 12 initial condition of material at risk within the fuel 13 itself.

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

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

29 structural material?

1 DR. DENMAN: So the Structures Topical 2

Report will ensure that the vessel is not degraded 3

with -- by the FLiBe. Within our anticipated 4

operational range any other structure system and 5

component is not safety-related, and breaks in those 6

systems would be evaluated in our postulated event 7

analysis.

8 MEMBER REMPE: Okay. Thank you.

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

11 MEMBER REMPE: Okay. Thank you.

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

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

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

30 CHAIR PETTI: Okay.

1 DR. DENMAN: Okay. So and then there's 2

also going to be tritium that's going to be stored in 3

the graphite pebbles and structures that can be 4

desorbed at elevated temperatures and our methodology 5

will examine that release.

6 Our design-basis accident site boundary 7

dose will be used -- dose is going to demonstrate that 8

the KP-FHR meets dose limits in 10 C.F.R. 50.34, 9

52.79, and 100.11. Again, technical specifications 10 will be set on the activity of the FLiBe, cover gas 11 and other systems, and the system will be design to 12 preclude incremental fuel failures from DBA conditions 13 as evaluated by KP-BISON.

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

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

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

31 Our radionuclide grouping and transport 1

approach is very similar to that used in light water 2

reactor safety analysis. I have the MELCOR grouping 3

structure on the right here where you can see the 4

various chemical groups and then the representative 5

element at the top that represents now the releases 6

from those groups are calculated for light water 7

reactor. We take a similar approach, although we 8

evaluate the grouping structures specifically to the 9

barrier and the release mode within that barrier.

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

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

32 into the gas space are evaluated with RADTRAD and 1

ARCON.

2 CHAIR PETTI: So, Matthew, I had bad 3

network quality there for a minute so I missed it, but 4

the groupings are the same no matter where the fission 5

product is in the system, or does it -- when it's in 6

the fuel it's considered one way because of the 7

chemistry there. When it's in the salt it's 8

considered because of the chemistry there?

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

16 CHAIR PETTI: Okay.

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

33 Advanced Gas Reactor Development Program as mentioned 1

earlier.

2 Our TRISO fuel manufacturing 3

specifications will determine how the fuel 4

configurations begin in the transient. Fission 5

products will diffuse from imperfect particles, 6

primarily particles with failed silicon carbon layers 7

or that have exposed kernels. Radionuclides from 8

heavy metal contamination from the manufacturing 9

process will have no credit for radionuclide retention 10 in steady state.

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

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

18 For the FLiBe barrier this is the second 19 part of our functional containment for radionuclide 20 retention. Once radionuclides are in the FLiBe they 21 will be separated into either salt soluble compounds, 22 suspended oxides, noble metals or gases. And Kairos 23 Power's Fuel Development Program, or sorry, FLiBe 24 Development Program builds on radionuclide retention 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

34 experience in the molten salt reactor experiment with 1

the exception that our salt is going to be much, much 2

cleaner than what was experienced in MSRE, which was 3

a fuel salt system.

4 CHAIR PETTI: So, Matthew, just a question 5

on that, and if I get into proprietary stuff, just 6

tell me and we'll cover it in the closed session.

7 I noticed that you had put some fissile 8

impurities in the salt, and I was surprised at that 9

level being that high. And I wasn't sure if that was 10 just being conservative or what was done back in the 11 old days of MSRE or whether that actually is what you 12 get.

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

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

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

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

35 CHAIR PETTI: So that's sort of a -- let's 1

just say a conservative level. That may not be what 2

you actually see in practice.

3 DR. DENMAN: Correct.

4 CHAIR PETTI: Okay.

5 DR. DENMAN: For tritium transport, 6

tritium transport within structures is determined by 7

mass transfer coefficients from FLiBe flow 8

characteristics throughout the system. Transport 9

within structures is determined based upon material 10 properties such as diffusion within and through steel, 11 diffusion and trapping within pebbles and structural 12 graphite.

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

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

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

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

36 It is now understood better that simple 1

experiments where one injects tritium into molten 2

coolants like this show a bias. The tritium doesn't 3

actually go in. It can sit along the surface. If you 4

think of like a loop. This has been shown in Europe 5

in the Fusion Program for a different coolant that's 6

a low-solubility coolant. FLiBe is a low-solubility 7

coolant.

8 And I think it's very difficult. This is 9

exactly how I would model it. I just think the 10 validation is going to be quite challenging because 11 the experiments may have these biases that you --

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

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

37 very different microstructures so that there's very 1

big differences potentially in trapped concentrations.

2 I believe the trapped energies are probably generic to 3

carbon materials, but the actual concentrations are 4

very strongly microstructural-dependent. Radiation 5

can affect it, too. All of these things make it much 6

more complicated than these really nice elegant 7

models.

8 And if you go back -- you have to go back 9

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

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

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

38 concern. So it's worth looking at some of that as you 1

think about how you're going to bound things and do 2

sensitivity analysis and the like.

3 DR. DENMAN: Thank you very much for your 4

feedback on that. It's very valuable and insightful 5

and we'll take it as we move forward with this 6

approach.

7 For tritium from the FLiBe-free surface, 8

tritium fluoride and tritium gas can both exist as 9

dissolved gases in the FLiBe. Contributions to off-10 site dose would either require permeation into vessel 11 or piping and then release into the reactor building 12 or evolution into the gas space which is modeled via 13 the gas transport equations influenced by the 14 experiments as shown below.

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

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

39 transient and use the Henry correlation for aerosol 1

settling, and conservatively and prescriptively 2

leakage rates will be applied out of the reactor 3

building into the environment.

4 For ARCON the release definitions around 5

the release of radioactive material from the site the 6

location of the receptor and meteorological conditions 7

at the site are needed to calculate chi over qs.

8 Various limitations are set forth in this 9

topical report. They are listed on the slide, but I 10 will not read them word for word.

11 And with that are there any further 12 questions?

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

40 deal with it are a problem. So it's just something to 1

put on your tickler list as your design evolves.

2 Again my information may be a little out 3

of date, but this was at least the case ten years ago, 4

but they were still struggling with some of these 5

sorts of issues.

6 DR. DENMAN: Thank you very much for that 7

feedback. It's definitely something that we'll take 8

back as we continue to mature our design.

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

11 CHAIR PETTI: Yes, members, any questions?

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

16 CHAIR PETTI: Okay. Thanks.

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

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

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

23 MR. WANG: Dave?

24 CHAIR PETTI: Yes.

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

1 CHAIR PETTI: Yes, yes, yes. So okay, 2

let's open -- anybody that has a comment from the 3

public, *6 to un-mute yourself. Give us your name and 4

your comment.

5 Okay. Hearing none, I guess we will end 6

this open session. And I think all the members should 7

have the link to the closed session.

8 And, Kairos, we'll want you to make sure 9

that all the folks you think should be there should be 10 there and Weidong and our staff will handle the NRC 11 side.

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

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

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

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

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

KP-FHR Mechanistic Source Term Methodology Topical Report ACRS Subcommittee Meeting, November 19, 2021

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 affordableand safe.

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

3 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).

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

4 Sources of Steady State Material at Risk (MAR)

Fuel Assumed to retain FPs and HM Intact Layers Compromised IPyC Layer Compromised OPyC Layer Assumed to release FPs Compromised SiC Layer Exposed Kernel Inservice Failures No Retention of FPs or HM Dispersed Uranium Structures (Graphite + Pebbles)

Tritium C.A. Contamination Circulating Activity Initial Salt Loading

Uranium, Thorium, Metals

FP and HMC from Fuel Transmutation and Fission Tritium Production Offgas Gases and Vapors Building Tritium Nitrate Tritium Flibe Cleanup Noble Metals

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

5 Sources of Steady State Material at Risk (MAR)

Fuel Assumed to retain FPs and HM Intact Layers Compromised IPyC Layer Compromised OPyC Layer Assumed to release FPs Compromised SiC Layer Exposed Kernel Inservice Failures No Retention of FPs or HM Dispersed Uranium Structures (Graphite + Pebbles)

Tritium C.A. Contamination Circulating Activity Initial Salt Loading

Uranium, Thorium, Metals

FP and HMC from Fuel Transmutation and Fission Tritium Production Offgas Gases and Vapors Building Tritium Nitrate Tritium Flibe Cleanup Noble Metals Fuel Manufacturing Specifications

+

KP-BISON (Fuel Performance Topical)

+

Burnup (Core Design Topical)

Circulating Activity Technical Specifications Structures (Graphite + Pebbles)

Tritium Source Term Methodology Flibe Cleanup Technical Specifications Offgas Technical Specifications Building Technical Specifications Nitrate Technical Specifications

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6 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

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

7 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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8 MAR Mobilization in AOOs, DBEs, and 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.

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

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

9 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.

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

10 Radionuclide Grouping and Transport Approach

Transport of radionuclides through each 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.

LWR Example

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11 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

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

12 TRISO Fuel Configurations for Material at Risk (MAR)

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

Fission products (FPs) diffusion from imperfect particles Compromised SiC layer but intact IPyC and/or OPyC will release a smaller a range of FPs to the Flibe.

Exposed kernels and in-service TRISO failures mobilize mobile a larger range FPs transported to the Flibe at steady state.

Radionuclides from heavy metal contamination from the manufacturing process will have no credited retention at steady state.

Minimal expected steady state diffusion of radionuclides are expected from the remaining configurations.

The compromised configurations will be partially or entirely depleted of FPs during steady state, thus reducing the MAR available for release during the transient.

Fuel Assumed to retain FPs and HM Intact Layers Compromised IPyC Layer Compromised OPyC Layer Assumed to release FPs Compromised SiC Layer Exposed Kernel Inservice Failures No Retention of FPs or HM Dispersed Uranium TRISO Cohorts

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

13 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 gas phases

Kairos Powers Flibe development program builds on radionuclide retention experience in the Molten Salt Reactor Experiment Impurities

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

14 Tritium Uptake into Structures Metallics, Fuel and Moderator Pebbles, Structural Graphite Heat Exchanger - Permeation through Metal Core - Retention in Pebbles Downcomer - Vessel Permeation and Reflector Retention

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:

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15 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.

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

16 Gas Space Analysis Simple and Conservative

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

Key Inputs:

RADTRAD:

Mobilized material-at-risk activities Depletion mechanisms Radioactive decay and/or Henry correlation for aerosol settling.

Leakage rates (Conservative)

ARCON:

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17 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.

ML21356A607

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