ML23137A037
| ML23137A037 | |
| Person / Time | |
|---|---|
| Issue date: | 05/02/2023 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-2387 | |
| Download: ML23137A037 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards General Atomics Licensing Subcommittee Docket Number:
(n/a)
Location:
teleconference Date:
Tuesday, May 2, 2023 Work Order No.:
NRC-2387 Pages 1-109 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
(ACRS) 5
+ + + + +
6 GENERAL ATOMICS DESIGN SUBCOMMITTEE 7
+ + + + +
8 TUESDAY 9
MAY 2, 2023 10
+ + + + +
11 The Subcommittee met via Teleconference, 12 at 12:30 p.m. EDT, Vicki M. Bier, Chair, presiding.
13 14 COMMITTEE MEMBERS:
15 VICKI M. BIER, Chair 16 RONALD G. BALLINGER, Member 17 CHARLES H. BROWN, JR., Member 18 VESNA B. DIMITRIJEVIC, Member 19 GREGORY H. HALNON, Member 20 WALTER L. KIRCHNER, Member 21 JOSE MARCH-LEUBA, Member 22 DAVID A. PETTI, 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 ACRS CONSULTANTS:
1 DENNIS BLEY 2
4 DESIGNATED FEDERAL OFFICIAL:
5 WEIDONG WANG 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 AGENDA 1
Item Page 2
ACRS Chairman Introductory Remarks 4
3 NRC Staff Introductory Remarks 7
4 GA-EMS Introductory Remarks...........
9 5
Overview of GA-EMS FMR Design.......... 11 6
FMR Principal Design Criteria Topical 7
Report - GA-EMS.............. 56 8
Draft Safety Evaluation for the FMR Principal Design 9
Criteria Topical Report - NRC Staff.... 68 10 Public Comments................
108 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 P-R-O-C-E-E-D-I-N-G-S 1
12:30 p.m.
2 CHAIR BIER: This meeting will now come to 3
order. This is a meeting of the General Atomics 4
Licensing Subcommittee of the Advisory Committee on 5
Reactor Safeguards.
6 I am Vicki Bier, chairman of today's 7
Subcommittee meeting. Here as members in attendance 8
are David Petti, Charles Brown. Jose is here. Joy 9
Rempe, Matt Sunseri, Ron Ballinger. Walt Kirchner I 10 think will be back in a minute, probably. Greg Halnon 11 is here.
12 Vesna, are you online? I can't really 13 see.
14 MEMBER DIMITRIJEVIC: Yes, I am.
15 CHAIR BIER: Yes.
16 MEMBER DIMITRIJEVIC: Hi.
17 CHAIR BIER: Great. Thank you. And how 18 about our consultants, Dennis Bley and Steve Schultz?
19 DR. BLEY: Dennis here.
20 CHAIR BIER: And it looks like Steve is 21 also here. Apologies. I have to keep taking my 22 glasses on and off for different distances. Okay.
23 Weidong Wang of the ACRS staff is the 24 Designated Federal Official for this meeting.
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 During today's meeting the Subcommittee 1
will review the staff's draft safety evaluation on the 2
General Atomics Fast Modular Reactor principal design 3
criteria. The Subcommittee will hear presentations by 4
and hold discussions with the NRC staff, General 5
Atomics' representatives, and other interested persons 6
regarding this matter.
7 Parts of the presentations by the 8
applicant and the NRC staff may be closed in order to 9
discuss information that is proprietary to the 10 licensees and its contractors pursuant to 5 USC 552 11 (b)(C)(iv).
12 Attendance in the meeting that deals with 13 such information will be limited to the NRC staff and 14 its consultants, General
- Atomics, and those 15 individuals and organizations who have entered into an 16 appropriate confidentiality agreement with them.
17 Consequently, we will need to confirm that we have 18 only eligible observers and participants in any closed 19 part of today's meeting.
20 The rules for participation in all ACRS 21 meetings, including today's, were announced in the 22 Federal Register on June 13, 2019.
23 The ACRS was established by the Atomic 24 Energy Act and is governed by the Federal Advisory 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 Committee Act.
1 For background, the ACRS is intended to be 2
independent of the NRC staff. ACRS issues publicly 3
available letter reports that provide the Commission 4
our independent technical reviews of NRC staff 5
evaluations of the safety of proposed reactor 6
facilities.
7 It is required by the Atomic Energy Act 8
that ACRS participate in the reviews of submittals for 9
new reactor licenses. As part of our review, we 10 consider not only the staff's safety evaluations but 11 also the original submittals by the applicant.
12 As part of our review process, ACRS 13 members will ask questions and at times make 14 statements. However, these statements are individual 15 member opinions and should not be construed as ACRS 16 findings or opinions. ACRS opinions are only as 17 documented in our written letter reports.
18 The ACRS section of the U.S. NRC public 19 website provides our charters, bylaws, agendas, letter 20 reports, and full transcripts of all full and 21 subcommittee meetings, including the slides presented.
22 The meeting notice and agenda for this meeting were 23 also posted there.
24 So far we have received no written 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 statements or requests to make an oral statement from 1
members of the public.
2 The Subcommittee will gather information, 3
analyze relevant issues and facts, and formulate 4
proposed positions and actions as appropriate for 5
deliberation by the full Committee.
6 A transcript of today's meeting is being 7
kept and will be made available.
8 Today's meeting is being held in person 9
and over Microsoft Teams for ACRS staff and members, 10 NRC staff, and the applicant. There is also a 11 telephone bridge line and a Microsoft Teams link 12 allowing participation of the public.
13 When addressing the Subcommittee, 14 participants should first identify themselves and 15 speak with sufficient clarity and volume so that they 16 may be readily heard. When not speaking, we request 17 that participants mute your computer microphone or 18 phone by pressing star 6.
19 We will now proceed with the meeting. And 20 I'd like to start by calling up the NRR staff. And I 21 believe that will be, sorry, Candace De Messieres.
22 Sorry if I mispronounced that. Thank you.
23 MS. DE MESSIERES: Thank you, Chair Rempe 24 and also Member Bier for the opportunity to present 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
8 the Committee today.
1 So I'm Candace De Messieres, Chief of the 2
Advanced Reactor Technical Branch 2 in the Division of 3
Advanced Reactors and Non-Power Production and 4
Utilization Facilities, or DANU, in the Office of 5
Nuclear Reactor Regulation.
6 Later in this meeting after the General 7
Atomics design overview, the NRC staff will provide 8
you with a summary of our review of the General 9
Atomics Electromagnetic Systems, or GA-EMS, Fast 10 Modular Reactor Principal Design Criteria Topical 11 Report.
12 Like the light water-based general design 13 criteria contained in Part 50, Appendix A, the PDC 14 established the necessary
- design, fabrication, 15 construction, testing, and performance requirements 16 for structures, systems, and components that are 17 important to safety. Accordingly, generation of 18 adequate PDC is a foundational step on the path to 19 licensing.
20 In our review of the GA-EMS PDC Topical 21 Report, the NRC staff leveraged the information in 22 Regulatory Guide 1.232 that was reviewed by the ACRS 23 in 2018 and provides guidance for developing generic 24 advanced reactor design criteria, or ARDC, 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
9 technology specific sodium-cooled fast reactor and 1
modular high temperature gas-cooled reactor PDC.
2 The staff also drew on its experience 3
reviewing PDC for other advanced non-light water 4
reactors, such as the Kairos Power fluoride salt-5 cooled high temperature reactor.
6 Thank you again for the opportunity to 7
present to the Committee. And we look forward to 8
hearing your insights and feedback later in the 9
meeting. Thank you.
10 CHAIR BIER: Okay. I believe it is now 11 time for the General Atomics introductory remarks by 12
-- I'm not sure if that's -- oh, sorry, that's Aaron 13 Majors I believe. And I don't know if you're in the 14 room or online.
15 MR. MAJORS: I am online. Everyone hear 16 me clearly?
17 CHAIR BIER: Yes.
18 MR. MAJORS: Thank you so much. I just 19 want to start by saying thank you for taking the time 20 out to have this review, very needed. And we're 21 looking forward to hearing from the outcome of this 22 meeting.
23 I'd like to say just a couple quotes that 24 are apropos for safety. These authors are unknown.
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
10 But no safety, know pain, but if you know safety, you 1
have no pain. Also, safety doesn't happen by 2
accident.
3 And those are the safety moments that we 4
live by here at General Atomics. General Atomics has 5
a long history of developing extremely safe reactors.
6 And this history began with the TRIGA research in 7
reactors and has evolved into high temperature gas-8 cooled reactors is where we are today with our Fast 9
Modular Reactor, which is an answer to a growing 10 market and the need for small, easily deployable 11 reactors that provide great stability through rapid 12 load following.
13 And so our main objective is the 14 achievement of proper operating conditions and the 15 prevention or mitigation of accident consequences to 16 protect our workers, the public, and the environment 17 from radiation hazards.
18 So we're really happy to be here and 19 looking forward to the outcome. Thank you.
20 CHAIR BIER: Okay. Thank you. We're 21 happy to have you here.
22 So now the first part of the presentation 23 is the overview of the General Atomics Fast Modular 24 Reactor design. And I believe the presenter for 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
11 is going to be John Bolin. Is that correct?
1 MR. BOLIN: That's correct.
2 CHAIR BIER: Excellent. Welcome.
3 MR. BOLIN: Okay. Let's see. Is this 4
displaying as just a single slide?
5 CHAIR BIER: Yes. We now see your slides.
6 MR. BOLIN: Okay. All right. So this is 7
going to be an overview of the conceptual design to 8
date. And I'll continue. And I am the safety and 9
licensing lead here at GA-EMS for the Fast Modular 10 Reactor.
11 So, before I go on to this goal, I wanted 12 to introduce our team. We have a very distinguished 13 team of collaborators, including a
strategic 14 partnership with Framatome, on this Fast Modular 15 Reactor design. We have worked with Framatome in the 16 past on the gas turbine-modular helium reactor and 17 worked together and competed against each other on the 18 next generation nuclear plant.
19 We also have on our team EPRI. And EPRI 20 has, as part of their team, they have enlisted to help 21 Vanderbilt University.
22 We also have two other universities that 23 are collaborating with us, the University of 24 Wisconsin-Madison, under the leadership of Mike 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 Corradini, and the University of Texas at Arlington.
1 And they're focused on the turbine machine design.
2 We also have the expertise of three 3
national labs, Idaho National Lab, with both their 4
BISON and ATR and TREAT expertise. And we also have 5
Argonne National Lab, with their fast reactor fuel 6
design expertise, and Sandia National Lab, with their 7
MELCOR modeling expertise.
8 So the goal is to develop a Fast Modular 9
Reactor. It's 44 megawatts electric. And, you know, 10 it's intended for flexible power generation and easily 11 dispatchable and carbon free. And we're targeting 12 commercial operations by 2035.
13 The team is developing key design 14 attributes. It is a fast spectrum reactor. We use 15 helium inert gas as coolant. We have pellet loaded 16 fuel rods. We are emphasizing site flexibility and 17 small passive heat removal systems that will result in 18 safe, maintainable, and cost effective nuclear power 19 generation.
20 The FMR project officially started on 15th 21 of December of 2021. It's a three-year program under 22 the ARDP, Advanced Reactor Concepts 2020 Program.
23 MEMBER REMPE: Hey, John. This is Joy.
24 You know how ACRS members always are rude 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
13 interrupt people. I don't know if --
1 MR. BOLIN: And I was going to say you 2
guys can ask me questions any time.
3 MEMBER REMPE: Things haven't changed over 4
years. But anyway, I don't know if you're too close 5
to the mic or there's some tapping sounds. But it's 6
hard to sometimes hear what you're saying. Do you 7
have an idea what it could be? And maybe --
8 MR. BOLIN: It might be my coffee flask is 9
jiggling a little bit. So maybe that's --
10 MEMBER REMPE: Okay. That would help.
11 Thank you. Sorry to interrupt. But it was getting 12 distracting. Thanks.
13 MR. BOLIN: Okay. All right. We'll see 14 if that's improved.
15 MEMBER REMPE: That is better. Thanks.
16 MR. BOLIN: Okay. I'll go on to the Next 17 slide. So the project objectives, you know, their 18 focus is to enable future deployment, development and 19 deployment. And so we're particularly interested in 20 verification of key metrics in fuel, safety, and 21 operational performance.
22 So, as stated here, we will look at the 23 technical feasibilities. I mean, basically the 24 conceptual design effort is to prove the technical 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 feasibility of the design, looking at high burn-up 1
fuel operation, passive safety features, and rapid 2
grid adaptability or load following.
3 The project obviously includes pre-4 application licensing activities with the NRC. That 5
was a key desire of the DOE in their FOA. And the 6
project will also conclude with an initial cost 7
evaluation.
8 Like I said, the two focuses are on 9
verification, both experimental and numerical 10 verification. The experimental verification, we do 11 have a fuel fabrication campaign that will result in 12 a high burn-up irradiation test at, and transient test 13 at ATR and TREAT to begin the qualification of the 14 fuel design. And we'll go into that a little more in 15 the later slides.
16 We also have scaled tests of the reactor 17 vessel cooling system using the facility that 18 University of Wisconsin-Madison has to further verify 19 the passive cooling capability. In this case, the 20 RVCS test facility that they have is actually between 21 half scale and full scale of our RVCS design. So it 22 will be a very interesting test.
23 We're also doing numerical verification.
24 Part of this is the accident analysis work being done 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
15 both, or being done at UW-M. They are developing a 1
MELCOR model. So that will support the design work 2
and pre-application licensing.
3 We are also doing, with Sandia, a MELCOR 4
model to simulate the FMR plant and to demonstrate 5
rapid load following capability, also load rejection 6
and basically a variety of operational transients.
7 Any questions so far? Okay.
8 Okay. This is the, this goes over our 9
effort to design the FMR core to improve safety 10 margin. Some of the things to note on this slide is 11 the core power density. Oh, and I should -- and so, 12 in this slide, I'm comparing numbers for the Fast 13 Modular Reactor, the gas turbine-modular helium 14 reactor, also designed by General Atomics, and the 15 AP1000 PWR.
16 So, I mean, the first state, of course, is 17 the output. The reactor output is quite low. It's 18 100 megawatts thermal, you know, 6 times lower than 19 the GT-MHR and much lower than the AP1000.
20 The power density is almost 15 megawatts 21 per meter cubed. That's higher than the GT-MHR but 22 much less than the AP1000.
23 The heat generated in the fuel, actually 24 our number is similar to AP1000. Most of the heat 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
16 does get deposited in the fuel, of course. Pressure 1
is 7 megapascals.
2 The other thing to note, it's not 3
explicitly mentioned here. So the outlet temperature 4
is not, you have to calculate the outlet temperature 5
based on these numbers. So the FMR has an outlet 6
temperature of 800 degrees C, while the GT-MHR had an 7
outlet temperature of 850 degrees C. So we cut back 8
the outlet temperature a little bit to improve safety 9
margin.
10 The other thing to note, of course, is 11 the, similar to the power density, the fuel rod 12 average linear power is quite low, much lower than the 13 AP1000.
14 And the other thing is the fuel height.
15 DR. BLEY: John?
16 MR. BOLIN: Yes.
17 DR. BLEY: This is Dennis Bley.
18 MR. BOLIN: Yes, Dennis.
19 DR. BLEY: I'm remembering back a long 20 time ago, probably from the '70s. Excuse me. You had 21 a fast reactor design way back then. And if you lost 22 force circulation, you had about 45 seconds I think, 23 if my memory is right, to get it back to prevent 24 significant damage. How does this reactor look if you 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
17 lose force circulation?
1 MR. BOLIN: Well, there's two things that 2
are in our favor. First off, we are using SiGA 3
silicon carbide composite cladding. In fact, the 4
whole fuel assembly is a ceramic composite cladding or 5
ceramic composite material. And so it has a much 6
higher temperature capability. But also we have 7
greatly reduced the power density compared to the, I 8
think you're referring to the gas-cooled fast breeder 9
reactor --
10 DR. BLEY: That's probably true.
11 MR. BOLIN: -- back in the '70s. So the 12 power density is much less.
13 And so, while I'm not going to present the 14 accident results, the passive safety, we have 15 engineered that so that we can safely cope with a loss 16 of force circulation, loss of force cooling.
17 DR. BLEY: Okay. Thanks. I look forward 18 to seeing more about that later.
19 MR. BOLIN: Sure. And like I said, so the 20 cladding material is a SiGA silicon carbine composite.
21 And we'll go into that a little bit more.
22 And the core height is also quite small 23 compared to the other two designs shown here. It's 24 only 1.8 meters in height. And this is the active 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 height. This is the fuel zone of the fuel assembly.
1 Okay. Any questions before I move on?
2 MEMBER PETTI: John, this is Dave Petti.
3 Are you going --
4 MR. BOLIN: Hi, Dave.
5 MEMBER PETTI: Hi. Are you going to show 6
us some pictures of what a fuel assembly looks like?
7 MR. BOLIN: Yes, definitely.
8 MEMBER PETTI: Okay. Then I will wait.
9 Thanks.
10 MR. BOLIN: Okay. In fact, it's, part of 11 it is on the next slide here.
12 So the fuel design, it leverages both UO2 13 legacy fuel development and SiGA cladding development.
14 So we purposefully chose high density UO2 that's been 15 proven in LWRs and tested in fast reactors in order to 16 minimize the fuel development timeframe.
17 The silicon carbine composite cladding, 18 SiGA, it's undergoing testing and maturation to the 19 DOE accident tolerant fuel program. And in fact, SiGA 20 cladding is being irradiated presently in ATR.
21 The fuel design uses, actually uses the 22 ATF-LWR dimensions, you know, so that the cladding is 23 the same size as that being developed for ATF. But it 24 does have, the fuel design does have a large plenum 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 similar to what you find in the legacy liquid metal 1
fast reactor fuel designs --
2 MEMBER REMPE: John. Oh, I'm sorry. Go 3
ahead.
4 MR. BOLIN: Yes.
5 MEMBER REMPE: Okay. Did you finish your 6
last sentence? I didn't mean to cut you off.
7 MR. BOLIN: Yes. Go ahead.
8 MEMBER REMPE: Okay. Well, I was curious 9
if you could talk a little bit more about the end cap 10 welding. There's an image shown here (audio 11 interference) an end cap on with this SiGA material.
12 And apparently you've made it through leak testing and 13 pressure testing.
14 And how long has it been in the ATR? And 15 how long is it scheduled to be in the ATR? Are they 16
-- is it going through any PALM cycles in the ATR so 17 it's sort of having some ramp testing?
18 MR. BOLIN: I don't know the details of 19 that. I mean, we have made a lot of progress in the 20 end cap welding. And these are sealed rodlets that 21 have been hermetically tested and meet the hermeticity 22 requirements. And they will go through a few cycles 23 I believe. I don't know if it will go through a PALM 24 or not. And I don't know the details of that.
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20 MR. MAJORS: It's six cycles that our 1
specimens will be in ATR.
2 MR. BOLIN: This is, I think she's 3
particularly, Joy is particularly asking about --
4 that's -- and Aaron is correct. So the ATF, I don't 5
know about the ATF cladding that's being irradiated.
6 The FMR cladding will also be, go through, like Aaron 7
said, it will go through up to six cycles.
8 MEMBER REMPE: But you've not started that 9
test yet --
10 MR. BOLIN: That hasn't started yet.
11 MEMBER REMPE: Okay.
12 MR. BOLIN: I'm going to --
13 MR. MAJORS: It starts in December.
14 MEMBER REMPE: Oh, okay. So it starts 15 this December. And what is the peak temperature that 16 this end cap weld has survived to date?
17 MR. BOLIN: I don't know the answer to 18 that.
19 MEMBER REMPE: Okay. I just am curious.
20 I mean, it's not necessary for this PDR report, but, 21 or PDC report, but I just am --
22 MR. BOLIN: Yes.
23 MEMBER REMPE: -- curious on how far, 24 because I know that was an issue for a lot of years.
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21 MEMBER PETTI: John, just a question on 1
the diameter of the UO2, because you mentioned about 2
fast and thermal. Is it the size of a thermal UO2 or 3
a fast reactor UO2 or somewhere in between?
4 MR. BOLIN: Well, I believe the fast 5
reactor UO2 was extremely small.
6 MEMBER PETTI: Right.
7 MR. BOLIN: So it's not like that. But --
8 MEMBER PETTI: Okay.
9 MR. BOLIN: -- density obviously is much 10 less in the liquid metal fast reactor similarly. So 11 the UO2 pellet diameter is a little bit smaller than 12 a standard UO2 pellet, because we do have a somewhat 13 larger gap between the pellet and the cladding --
14 MEMBER KIRCHNER: John, this is Walt 15 Kirchner.
16 MR. BOLIN: It's basically the same as an 17 LWR pellet.
18 MEMBER KIRCHNER: John, this is Walt 19 Kirchner. So I'm thinking back to the prior work that 20 GA did in this particular area. If I remember 21 correctly, you were looking at uranium carbide pellets 22 or platelets or different --
23 MR. BOLIN: Correct.
24 MEMBER KIRCHNER: -- designs, not UO2.
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 For this reactor with a longer lifetime, what is the, 1
what's the effective full power years of the UO2 in 2
terms of burn-up?
3 MR. BOLIN: I think I will cover that.
4 But it's 100 megawatt days per --
5 MEMBER KIRCHNER: Metric ton?
6 MR. BOLIN: Yes.
7 MEMBER KIRCHNER: That's pretty high 8
compared to the UO2 that's used, because --
9 MR. BOLIN: Currently licensed, correct.
10 It's higher than what's currently licensed. Fast 11 reactor oxide fuel tests get to that burn-up and 12 higher. But --
13 MEMBER KIRCHNER: Doesn't it center quite 14 a bit? I thought that's why you were looking at 15 uranium carbide and not UO2 previously.
16 MR. BOLIN: Well, the reason we were 17 looking at uranium carbide, and we still are pursuing 18 that reactor design, the centering is going to be much 19 lower than you might expect because the UO2, peak UO2 20 fuel temperature is much lower than LWRs. So, and 21 I'll show you that. I think I show you that later.
22 I might be getting my presentations mixed up.
23 But, you know, it's probably about, well, 24 no, we'll see that, about 1,200 degrees C is the peak 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 fuel temperature.
1 MEMBER KIRCHNER: And then the grid 2
material for the X bundle is what?
3 MR. BOLIN: Silicon carbide composite.
4 MEMBER KIRCHNER: Silicon carbide as well.
5 MR. BOLIN: Yes.
6 MEMBER KIRCHNER: Thank you.
7 MR. BOLIN: And the support tube is also 8
silicon carbide. And you can see a picture of silicon 9
carbide composite cladding and in the X-ray tomography 10 of a cladding tube, and then as Joy mentioned, the end 11 cap welding, which we think we have perfected. So, 12 and it's ready for testing.
13 MEMBER PETTI: John, the grid plate is 14 also silicon carbide?
15 MR. BOLIN: Yes.
16 MEMBER PETTI: Thank you.
17 MR. BOLIN: Okay. This goes into more 18 detail, a little bit more detail of the different 19 steps we've gone through to prepare for the ATR and 20 TREAT irradiation capsules. Like I said, we've 21 enlisted Argonne's help in looking at the BISON fuel 22 model, looking at fission gas release and swelling.
23 And all those eventual fuel failure mechanisms are 24 also part of their modeling efforts.
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24 And so that's informed our fuel design and 1
analysis and, in particular, the analysis of these 2
test rodlets. We are looking at both standard size 3
rodlets and reduced diameter rodlets.
4 And like I said, we're going to do both 5
ATR irradiation for up to six cycles. And we've also 6
designed the rodlets with different size gaps to look 7
at performance, you know, both standard fuel rod 8
performance and performance where there would be 9
pellet clad interaction in possible failure. So 10 that's a
design into the analysis and the 11 experimentation. So --
12 MEMBER PETTI: So, John --
13 MR. BOLIN: Yes.
14 MEMBER PETTI: -- just a question on the 15 clad. If this gets proprietary, let me know. But, 16 you know, the last time I looked at SiC-SiC cladding 17 for ATF, there were some seminal papers out of Oak 18 Ridge that, given the delta T across the clad, you get 19 some pretty serious tensile stress built up because of 20 differential or irradiation swelling across it. I 21 would imagine the lower power density helps you with 22 that --
23 MR. BOLIN: Correct.
24 MEMBER PETTI: -- delta T. So, but, you 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 know, they've gone to things like liners and stuff.
1 This is just SiC-SiC, nothing special?
2 MR. BOLIN: It's the standard ATF silicon 3
carbine composite cladding. So it has the monolithic 4
outside layer and the composite woven silicon carbide 5
fiber, infiltrated silicon carbide in the inner layer.
6 And you're correct. Certainly the, you 7
know, our power density is much lower than light water 8
reactors. So the thermal gradients are much lower.
9 Also, operating at a higher temperature is actually, 10 for silicon carbide is actually a benefit, too. So 11 swelling --
12 MEMBER PETTI: Sure.
13 MR. BOLIN: -- swelling and irradiation 14 damage is less at higher temperatures. So we have 15 both of those factors in our favor.
16 MEMBER PETTI: Thanks.
17 MEMBER REMPE: Out of curiosity -- I'm 18 sorry. Is someone else -- do you want to go first?
19 DR. SCHULTZ: That was me, Steve, Joy.
20 You go ahead.
21 MEMBER REMPE: Oh, well, I was just --
22 DR. SCHULTZ: I'll come in next.
23 MEMBER REMPE: Okay. I was curious about 24 the instrumentation and what you're trying to validate 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 with these, or verify with these tests. Are you, is 1
it just temperature or are you -- and then post-2 irradiation examinations or are you going to try and 3
do any other type of measurements online that --
4 MR. BOLIN: No, no other --
5 MEMBER REMPE: -- tests?
6 MR. BOLIN: No other measurements online.
7 But we will look at, post-irradiation examination 8
we'll look at fuel physical changes and fission gas 9
release. So that will be looked at.
10 Actually, it's in that box right there.
11 The PIE will look at fission gas release and fuel and 12 cladding deformation. And particularly in the cases 13 where we have reduced gap between the fuel and the 14 cladding, you know, there's a possibility of cladding 15 fracture that also needs to be looked at.
16 MEMBER REMPE: Will you have temperature 17 instrumentation in the tests themselves?
18 MR. BOLIN: The fuel itself will not be 19 temperature monitored, no.
20 MEMBER REMPE: Okay. And just to caution 21 22 MR. BOLIN: At least not in the ATR 23 capsule. I don't believe it's in the ATR capsule.
24 The TREAT capsule may have instrumentation for 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
27 transient test.
1 MEMBER REMPE: In ATR sometimes small 2
geometry changes that are within specifications can 3
lead to interesting changes in temperatures that you 4
don't expect. It's just a caution.
5 Anyway, Steve, go ahead.
6 DR. SCHULTZ: My question was related, 7
John. And that is, in the testing, are you going to 8
achieve those temperatures that you anticipate in the, 9
for the reactor design parameters? The first 10 question.
11 MR. BOLIN: Yes. In fact, we will have 12 higher fuel temperatures than FMR will experience.
13 We'll have higher temperatures.
14 DR. SCHULTZ: Good. And for the six 15 cycles of operation, what burn-up do you expect to 16 achieve in the fuel test?
17 MR. BOLIN: We will get close to 100 18 megawatt days for burn-up.
19 DR. SCHULTZ: Good. Thank you.
20 MR. BOLIN: It's a very, unfortunately I 21 think, but it is a very accelerated test.
22 DR. SCHULTZ: It certainly appears that 23 way. Thank you.
24 MR. BOLIN: Yes.
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28 This just goes over the FMR test rodlets.
1 And like I said, they are being fabricated. Actually, 2
the rodlets have been fabricated. And they are going 3
to be loaded with UO2 pellets actually next week. And 4
then they'll be, the final end cap will be welded on 5
and shipped to INL for insertion into ATR at the end 6
of the year. So the fuel pellet processing is 7
basically standard UO2 fuel pellet processing.
8 The other steps are part of the silicon 9
carbide composite cladding fabrication. The silicon 10 carbide fiber is braided together, then infiltrated 11 with silicon carbide, and then both infiltration and 12 then deposition of an outside silicon carbide layer.
13 Pellets are then loaded. And then the final end cap 14 is sealed. So this is obviously a key accomplishment 15 of our conceptual design effort is to actually make 16 these fuel rods and to have them tested.
17 We were particularly, it was particularly 18 important to us to not just do a paper study on 19 conceptual design, but to actually do, like we 20 mentioned earlier, experimental verification of the 21 design.
22 The other, of course, part of our defense 23 in depth is the vessel system. We have a, the vessel 24 is sized, you know, for normal operation AOO 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 conditions. The design code, the ASME design code, of 1
course, is used. That is Section 3, Division 5, the 2
2021 edition.
3 Right now the thickness is adequate for 4
300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or 35 effective full power years based 5
upon the code. The code data does suggest very little 6
change out to, because of the temperatures that we're 7
at, very little change out to 60 years. And so a 8
future code revision should not have an impact on our 9
vessel design.
10 But we're also, one of the key problems 11 with gas-cooled reactors, helium gas-cooled reactors 12 is, of course, helium leakage. And so we pay 13 particular attention to using seal welds at all joints 14 to minimize helium leakage.
15 And another interesting thing is that a 16 lot of accidents and even load following, you know, 17 involve flow reductions. As we'll go over on the next 18 slide, we'll see, I'll discuss about the flow 19 reductions through normal operation. But all these 20 events, because we're using a Brayton cycle, the 21 pressure load on the vessel decreases during these 22 flow reductions.
23 MEMBER MARCH-LEUBA: John, this is Jose 24 March-Leuba. Just a layman question, you mentioned 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 earlier an exit gas temperature of 800 degrees C.
1 MR. BOLIN: Correct.
2 MEMBER MARCH-LEUBA: What materials are 3
you using for the hot leg and the vessel? And what 4
temperatures do they have to survive?
5 MR. BOLIN: Well, so, like the GT-MHR, we 6
do have a cross vessel that connects the reactor 7
vessel to the power conversion unit. And so it has, 8
the hot gas in on the inside of this cross vessel.
9 There's an insulated layer on the inside of this cross 10 vessel that then protects.
11 And then we have cold helium. Cold is a 12 relative term, you know. It's 509 degrees C on the 13 outside of this duct. And, you know, the layer that 14 connects to the cross vessel sees that 509 degrees C.
15 So all of the vessel materials are 509 degrees C or 16 lower.
17 MEMBER MARCH-LEUBA: The gas is the one 18 that has 7 megapascals. Somebody has to contain the 19 helium. I hope you have thought through this. I 20 don't know anything about this, but --
21 MR. BOLIN: It certainly is something we 22 have dealt with on numerous, particularly the GT-MHR 23 gas-cooled reactor design has looked at this 24 extensively.
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31 MEMBER MARCH-LEUBA: Yeah, and another 1
question I know even more about. You don't mention 2
anything about reactivity control. How do you plan to 3
control reactivity?
4 MR. BOLIN: We have both control rods, I 5
think boron carbide control rods and shutdown rods.
6 They will have also silicon carbide cladding.
7 MEMBER MARCH-LEUBA: But they're not shown 8
here in the picture, right? I don't see --
9 MR. BOLIN: No, no. Just the upper drive 10 mechanisms are shown there.
11 MEMBER MARCH-LEUBA: Yeah, one important 12 concern when you go for the final certification will 13 be, priming along is the control rod has to be a 14 design, has to have a design temperature that is 15 higher than the fuel. In other words, you should not 16 have an accident when you can meld a control rod and 17 leave the fuel intact, because that would be bad.
18 MR. BOLIN: Yes.
19 MEMBER MARCH-LEUBA: So you're saying your 20 design --
21 MR. BOLIN: That's why we are using 22 silicon carbide cladding for the control rods. I 23 don't know. Yeah.
24 MEMBER MARCH-LEUBA: And you said boron 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 carbide inside?
1 MR. BOLIN: Yes.
2 MEMBER MARCH-LEUBA: Okay.
3 MR. BOLIN: I guess my final comment was 4
that the conceptual design has been completed on the 5
reactor vessel internals. We're still working on the 6
details of the power conversion system, which I'm 7
going to discuss on the Next slide. And also this 8
shows the arrangement of neutron shields around the 9
core, of course. And there is a core shroud that 10 protects the vessel top head from the high temperature 11 gas exiting the core. So that's also an insulated 12 layer that protects the top head.
13 Cold helium coming into the reactor goes 14 all the way around the vessel and down the outside 15 core barrel and into the lower portion of the vessel 16 head and then up through the core.
17 MEMBER PETTI: John?
18 MR. BOLIN: Yes?
19 MEMBER PETTI: Just a question on your 20 outer reflector. Is it stainless steel like in sodium 21 systems? Or I know you guys had a design once with 22 beryllium carbide as an outer reflector.
23 MR. BOLIN: No, we're using -- zirconium 24 silicide is our reflector that's adjacent to the fuel.
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33 It is zirconium silicide. It's a product that we're 1
developing that we think it is a better reflector. I 2
mean, it's not as good as stainless steel, but, 3
fortunately, it has higher temperature capability.
4 And it helps to minimize, then, fuel rod peakings 5
along the reflector edge.
6 MEMBER KIRCHNER: And the upper and lower 7
reflectors, are they the same or --
8 MR. BOLIN: The reflector that is right 9
next to the fuel is always going to be zirconium 10 silicide. Now, the outside reflector, outside of the 11 zirconium silicide, we'll be using graphite, and I 12 think, also, on the bottom.
13 MEMBER KIRCHNER:
Upper and lower 14 reflectors are graphite?
15 MR. BOLIN: Well, below -- like I said, 16 there's always a zirconium silicide layer immediately 17 next to the fuel, the core. So, both the upper and 18 lower part is, first, zirconium silicide, and then, 19 graphite. In the core, I don't think that's the case, 20 but in the outer reflector that's the case. And the 21 lower reflector, that definitely is the case, yes.
22 MEMBER PETTI: But the inner reflector 23 has, like, an annulus of graphite?
24 MR. BOLIN: Yes -- no, no. The inner 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 reflector is always zirconium silicide.
1 MEMBER BALLINGER: This is Ron Ballinger.
2 I understand 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, but I don't 3
understand 540,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. You say, "Code revision."
4 Is this talking about Division 5 again?
5 MR. BOLIN: Yes, but, right now, the 6
material, 316 stainless steel, is only allowed up to 7
300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
8 MEMBER BALLINGER: Right.
9 MR. BOLIN: And so, future ASME Code 10 revision is intended to extend that to 540,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
11 MEMBER BALLINGER: And that's in process?
12 MR. BOLIN: Yes. And the data, of course, 13 already exist and shows very little change between 14 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and 540,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. So, it's not 15 expected to have any design impact.
16 MEMBER PETTI: What's the vessel material 17 again?
18 MR. BOLIN: 316 stainless steel.
19 MEMBER PETTI: Okay.
20 MR. BOLIN: Okay. So, the next slide goes 21 over a little bit on the power conversion system.
22 This is fairly standard for gas turbine design. It is 23 a direct Brayton cycle. I know it's maybe hard to 24 see. And I don't go into the details on the core 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 side.
1 So, exit temperature from the reactor goes 2
to the turbine, where, obviously, it goes down in 3
temperature and pressure. It goes through the 4
recuperator. So, the turbine outlet temperature is, 5
basically, directing the reactor inlet temperature.
6 So, there's a heat exchange between these two fluid 7
streams.
8 From the recuperator, it goes to a pre-9 cooler that cools the helium before going to the low 10 pressure compressor. And we have an intercooler in 11 this design. So, from the low pressure compressor, 12 you go to the intercooler. So, the heat that is added 13 during the compression process in the low pressure 14 compressor is removed by the intercooler, and that, 15 then, goes to the high pressure compressor, and then, 16 goes to the recuperator, and then, to the reactor.
17 So, this provides a high efficiency. And 18 like we said, it's the net -- the electrical output 19 from the generator is 44 megawatts electric.
20 The other interesting thing that enables 21 rapid load following is that we are using a GA 22 product, a current magnet motor generator. It is a 23 variable frequency generator. So, it can change its 24 rotational speed as needed. So, we can change 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
36 flow rate to the reactor by controlling the frequency 1
and speed of the current magnet motor generator.
2 It goes through an AC/DC-AC frequency 3
converter to get to the grid frequency. And, of 4
course, that goes to the grid.
5 And so, this combination allows us to 6
change the speed of the generator, change the speed of 7
the turbine machinery, and thereby change the flow 8
rate through the core. But, at the same time, while 9
we're doing that, we're maintaining the frequency 10 that's being fed to the grid.
11 And so, it promotes both rapid load 12 following -- we're shooting for 20 percent per minute 13 load following changes, but, also, it promotes grid 14 stability. So, those are two things that we see as 15 being important to the market in the future, and 16 particularly, as the electric market gets more and 17 more intermittent renewable energy sources.
18 MEMBER MARCH-LEUBA: And this is Jose 19 again.
20 With those power rates 20 percent per 21 minute, have you analyzed what happens to the UO2 22 pellets?
23 MR. BOLIN: So, that is going to be part 24 of our modeling that is still ongoing.
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37 MEMBER MARCH-LEUBA: Yes, because in the 1
life of the reactors, if you ramp up like that, you 2
will blow up the zirconium.
3 MR. BOLIN: Well, you know, the --
4 MEMBER MARCH-LEUBA: Yeah.
5 MR. BOLIN: Yeah. And, I mean, there's 6
two things that are in our favor: lower power density 7
and the silicon carbide is quite strong. And although 8
sometimes it's viewed as a negative, the fact that 9
it's not very ductile also means it keeps its shape, 10 even if the pressure inside is increasing.
11 MEMBER REMPE: John, I just assumed that 12 the TREAT test would encompass that. That's not going 13 to -- you're not going to try to run the presentient 14 (phonetic) test to such changes?
15 MR. BOLIN: The TREAT test is a 16 reactivity-initiated accident. So, it is doing that, 17 a reactivity-initiated accident simulation, not a load 18 following test.
19 MEMBER REMPE: But the ramp and the change 20 in power, wouldn't that --
21 MR. BOLIN: It should bound any reactivity 22
-- it should bound any load following change.
23 MEMBER REMPE: Yes, I would think it 24 would. Either that or -- it seems like you would want 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 to test that before you start doing load following in 1
the reactor.
2 MEMBER BALLINGER: This is a very 3
different paradigm. I mean, we've got to stop 4
thinking about zirconium cladding, because the 5
zirconium, that's what the load following problem was 6
for in light water reactors. It was for zirconium 7
cladding. It only takes 10 or 20 degrees C to crack 8
the UO2 at delta T. So, it's really the silicon 9
carbide that's got to take it, and it's very rigid 10 compared to zirconium, and there's no environment.
11 So, it's a different way of having to 12 think of it. It requires a heck of a lot more data 13 and experiments, but it's a very different fuel 14 system.
15 MEMBER REMPE: Yes, but I just think I 16 would rather test it out of the reactor.
17 MEMBER BALLINGER: I'll agree with you.
18 (Laughter.)
19 MEMBER REMPE: Yes, ahead of time.
20 MEMBER BALLINGER: Yes.
21 MEMBER REMPE: And I would find some way 22 of doing it.
23 MEMBER BALLINGER: Yes.
24 MEMBER REMPE: But it's not included 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
39 your plans?
1 Also, there's a lot of feedback, and I 2
don't know -- John, again, there's something with your 3
system, it looks like, according to the computers here 4
in the room.
5 MR. BOLIN: Oh, well, I mean, it could be 6
the fan in my laptop. So, I'm not sure I could 7
control that.
8 MEMBER REMPE: We did have that happen 9
with one of our members. Don't turn off your 10 computer. But, anyway, it started up recently, yes, 11 but anyway, you're going to have to have your managers 12 buy you a better computer.
13 (Laughter.)
14 MEMBER REMPE: But, anyway, yes, so you do 15 not have any plans to try and -- you're just going to 16 do it by analysis?
17 MR. BOLIN: Just do it by analysis for 18 now, yes.
19 CHAIR BIER: While we're paused here, I 20 also to check and see if Aaron had any comments or 21 clarifications that he wanted to add. Because 22 anything in the chat does not make it into the 23 transcript and the public meeting.
24 I guess maybe not right now.
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40 MEMBER KIRCHNER: So, John, this is Walt 1
Kirchner.
2 So, on this one, you're probably not going 3
to try and put a bottoming Rankine cycle on this?
4 MR. BOLIN: No.
5 MEMBER KIRCHNER: No? Okay.
6 MR. BOLIN: The temperature coming out the 7
recuperator is on the order of between 150 and 200 8
degrees C. So, there's not much left to take out.
9 MEMBER KIRCHNER: Okay. Right.
10 MR. MAJORS: Chairman? Chairman, my 11 apologies, I was on mute talking. I'm sorry.
12 CHAIR BIER: Okay, go ahead, Aaron.
13 MR. MAJORS: I was just following up.
14 Someone had asked the question about the peak 15 temperature survival for the end caps. And I just 16 wanted to add -- I didn't get a chance to interject 17 because you guys were rapid-firing questions, which is 18 great. We're trying to transcribe all these questions 19 as well for our technical team. But I just wanted to 20 add some points of reference.
21 So, our high-temperature gas application 22 for joining our end caps is sufficient for normal 23 operating and accident performance. So, our end caps 24 are stable up to 1900 degrees Celsius for inert 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
41 environments and 1750 for steam.
1 MEMBER REMPE: Thank you.
2 CHAIR BIER: Any other questions or 3
comments for Aaron?
4 MEMBER KIRCHNER: Could you share who your 5
fuel vendor is, or is that TBD?
6 MR. MAJORS: Fuel vendor is TBD.
7 MEMBER KIRCHNER: TBD? So, it's not one 8
of the fly rod manufacturers? I'm thinking of Orvis, 9
Sage, Winston. This is a joke. The technology that 10 you're describing is what's used to make graphite fly 11 rods for the cladding.
12 MEMBER BROWN: They'll use Loomis.
13 MEMBER KIRCHNER: Loomis? Okay.
14 (Laughter.)
15 MEMBER REMPE: And so, maybe this was in 16 the reading material that I've forgotten, but this 17 variable frequency generator, has anyone used it? Has 18 it been built? What's its status?
19 MR. BOLIN: This is a proven product at 20 GA-EMS. We have built an 8-megawatt version of both 21 the current magnet motor generator -- and the 22 frequency converter, we build that in modular fashion.
23 So, we have, I believe -- and, Aaron, you can correct 24 me -- I think we have 1-megawatt electric modules 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
42 we have built. And so, they're assembled in a modular 1
fashion.
2 And so, for the 44-megawatt, we will be 3
scaling both the current magnet motor generator up and 4
the frequency converter up. But we have already 5
proven the design at a reduced scale.
6 MEMBER REMPE: Okay. Thank you.
7 MR. BOLIN: And also on this slide, I 8
wanted to address -- we purposefully chose to use dry 9
cooling, even though that does have an efficiency 10 penalty. But that, clearly, will reduce the impact on 11 water resources and expands our siting options, 12 particularly, if you consider that a lot of the solar 13 and wind generation is going to be in possibly dry 14 areas of the West. So, that was a deliberate 15 selection on our part.
16 MR. FAIBISH: John, can I chime in? This 17 is Ron Faibish with General Atomics. Can I chime in 18 on something about silicon carbide?
19 MR. BOLIN: Sure.
20 MR. FAIBISH: I just wanted to add --
21 there were a lot of questions about the cladding. And 22 we have had campaigns at HFIR in Oak Ridge on 23 irradiation and prototypical conditions of 800 degrees 24 C of (audio interference) and outlets.
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43 So, variability was both shown -- and this 1
was, actually, a campaign back for EM-Squared back 2
when we were actively pursuing that design. The 3
outlet temperature of 800 degrees C is very 4
applicable, obviously, for FMR.
5 And also, in addition to that, as John 6
mentioned, there is a campaign starting at ATL to do 7
additional testings. So, that's up and coming.
8 But I just wanted you to know that silicon 9
carbide has been exposed to irradiation and to high-10 temperature conditions and showing good results. And 11 I think we're going to get more information to you, as 12 needed, from previous tests. So, I just wanted to 13 chime in on that.
14 Thanks, John.
15 MR. BOLIN: Thank you, Ron. Okay. Let's 16 move on. Well, let's see.
17 And, obviously, in a defense-in-depth, you 18 know, the third barrier we have is the containment.
19 Now, this is, unlike a lot of gas-cooled reactors, 20 this is, actually, a leak-tight containment. It is 21 below grade, like most gas-cooled reactors have been 22 below grade, but it is a leak-tight containment.
23 This was a deliberate selection to prove 24 safety and siting. Obviously, it also has an impact, 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
44 then, on fuel development and qualification.
1 Clearly, with TRISO fuel, there's a high 2
standard that TRISO fuel has to meet. And by having 3
a containment, we provide that extra defense for 4
potential fuel failures during severe accidents.
5 This also shows the arrangement of the 6
reactor. I don't know how I can get a pointer on 7
this. Let's see. There we go. Can you see that?
8 Yes, you can.
9 So, obviously -- maybe not obvious -- this 10 was the arrangement that was presented in the 11 proposal. So, it doesn't reflect concept design work 12 to date, but here is the reactor vessel. And around 13 the reactor vessel is the reactor vessel cooling 14 system, which we'll cover in the Next slide. And 15 then, in another compartment is the power conversion 16 unit, the power conversion vessel. We also have --
17 which will also be covered in the next slide -- the 18 maintenance cooling system. So, that is an active 19 forced cooling system that is provided in case the 20 power conversion system is unavailable. It is non-21 safety-related. Well, I'm getting ahead of myself.
22 And also, above --
23 MEMBER PETTI: John, functionally, it's 24 like the shutdown cooling system in a HTGR?
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45 MR. BOLIN: Correct. Correct.
1 MEMBER PETTI: Thanks.
2 MR. BOLIN: Functionally, it's like the 3
shutdown cooling system.
4 It is also like the direct reactor 5
auxiliary cooling system of liquid metal reactors, and 6
EM-Squared had that kind of a system, a forced cooling 7
system.
8 And up above here, we have RVCS water 9
tanks, which I'll also discuss in the Next slide. So, 10 this is just a general arrangement of the structure, 11 the below-grade structure, of the containment. And 12 it's a Category 1 structure.
13 The need for containment heat removal, 14
- cleanup, and
- venting, those are still under 15 investigation. It is something addressed as a 16 possibility in the PDCs as something that might be 17 necessary, but it's still under investigation.
18 Obviously, below-grade containment is 19 intended to also make us less vulnerable to airplane 20 crashes.
21 MEMBER PETTI: John, I wanted to go back 22 for a minute to the Brayton cycle.
23 MR. BOLIN: Go back to that slide?
24 MEMBER PETTI: No, no, just a question on 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
46 it.
1 On the bigger machines, there was always 2
a lot of development to get it to work, but at these 3
smaller sizes, I know that you can get such components 4
for gas systems. What's the status for helium 5
systems? Are they commercially available or?
6 MR. BOLIN: No, they're not commercially 7
available. We have not identified a manufacturer.
8 Clearly, there are a variety of turbine manufacturers 9
that we could choose from, but no one is building 10 helium turbine machinery. Obviously, there are air-11 driven --
12 MEMBER PETTI: Right.
13 MR. BOLIN: -- turbine machines, but no 14 helium ones.
15 MEMBER PETTI: Thanks.
16 MR. BOLIN: Sure. Okay. So, like I 17 alluded to, we have residual heat is being removed by 18 both active and passive systems. Really, the first 19 line of defense is the power conversion system itself.
20 We are, in particular, designing the system so that, 21 if there is a grid disruption, that the power 22 conversion system will ramp down and provide house 23 loads and cool the reactor at house loads.
24 But if the power diversion system is 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
47 source of the problem, then the maintenance cooling 1
system is available to remove core residual heat after 2
reactor shutdown. And so, like we said, it is similar 3
in function to the shutdown cooling system.
4 Basically, it's taking hot helium out of the reactor, 5
bringing it over helium-to-water heat exchanger, and 6
then, circulating that back into the reactor, and 7
basically, cooling the core. Details are still being 8
worked on.
9 And so, that water is cooled in a cooling 10 tower by forced air. So, all that system is intended 11 to be not safety-related.
12 The safety-related system is the RVCS, 13 although it has, actually, also has a non-safety-14 related component to it. So, the RVCS, the Reactor 15 Vessel Cooling System, has two loops. The panel of 16 two loops surrounding the reactor has alternating 17 tubes of one loop or the other loop.
18 The water in the RVCS circulates naturally 19 by buoyancy-driven flow. The water goes into a tank.
20 So, this tank -- there's two tanks. Like I say, 21 everything is redundant. There's also two of these 22 cooling towers, and both of these cooling towers have 23 a -- there's a heat exchanger in the RVCS tank that is 24 cooled by this cooling tower. So, it keeps the water 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
48 in the tank cold during normal operation -- both 1
tanks. Like I said, there's another cooling tower 2
with heat exchangers that's cooling this tank.
3 In an accident where we lose power, you 4
know, for the safety-related portion of the system, 5
the cooling tower is not safety-related. The pump for 6
this water is not safety-related. So, this whole 7
cooling system of the tank is not safety-related. So, 8
during an accident where we lose all non-safety-9 related systems, then the RVCS loses heat by boiling 10 off water from the water tank. And the water tank is 11 sized so that we have seven days of boil-off with just 12 one loop. And obviously, if we have both loops that 13 are functioning, then we have, you know, much longer 14 capacity to cool the reactor and vessel system.
15 Also, like many gas-cooled reactors, we 16 have an annular core arrangement that promotes passive 17 heat removal from the core through the reflector.
18 This, actually, like I said, this shows the zirconium 19 silicide reflector -- is this green. The blue is also 20 zirconium silicide. The central zone is also 21 zirconium silicide. And there is a graphite outer 22 layer outside of the zirconium silicide.
23 And there's three, yes, three fuel zones.
24 We have a three-batch core. And every 15 or more 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
49 years, we have a refueling and we replace one-third of 1
the core.
2 Let's see. I think that's all I need to 3
say about this slide. Any questions on this slide?
4 MEMBER KIRCHNER: So, John, this is Walt 5
Kirchner again.
6 MR. BOLIN: Yes?
7 MEMBER KIRCHNER: So, on this MCS, then, 8
you said that's not a safety-grade system. So, you 9
must have isolation valves to and from the helium 10 circuit?
11 MR. BOLIN: It is a --
12 MEMBER KIRCHNER: Because you have a water 13 heat exchanger there.
14 MR. BOLIN: We have isolation valves on 15 the waterlines, definitely. But the MCS is in the 16 containment. So, we have a --
17 MEMBER KIRCHNER: So, you're designing for 18 the potential that you would have a break in that line 19 20 MR. BOLIN: We'd have a break in --
21 MEMBER KIRCHNER: -- and the containment 22 would have the pressure rating --
23 MR. BOLIN: Correct.
24 MEMBER KIRCHNER: -- to withstand 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
50 blowdown? Okay.
1 MR. BOLIN: Correct. So, the MCS is one 2
of our sources of primary coolant breaks. Now, there 3
is a flow shutoff valve downstream -- it will probably 4
be downstream of the circulator. Because, otherwise, 5
we'll get natural circulation through this maintenance 6
cooling system. So, we have a flow shutoff valve for 7
normal operation.
8 MEMBER KIRCHNER: Thank you.
9 MR. BOLIN: Okay. This is my last slide.
10 Or, no, it's not my last slide. Okay.
11 So, one of the things that has been a 12 concern with gas-cooled reactors is bypass flow. And 13 also, related to that is flow-induced oscillations.
14 I know that there is a PDC on power oscillations.
15 The coolant itself, of course, it doesn't 16 have a reactivity effect, but movement of core and 17 reflector structures can have some reactivity effect.
18 In particular, the bypass flow, you know, because the 19
-- I don't know if it was clear, but the fuel assembly 20 is relatively open. So, it's kind of an open bundle.
21 It does have brackets on the corners, but, basically, 22 as the flow comes up through the fuel assembly, it can 23 redistribute among the various fuel assemblies.
24 But, also, because it's open, it can also 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
51 feed bypass flow paths. So, as you see here, around 1
the outer fuel assemblies, and also, all throughout 2
the inner fuel assemblies, there are gaps between 3
these blocks. And so, because of the open fuel 4
assembly, we can get bypass flow going from the fuel 5
area into the central zone and into the other 6
reflector zone. And, in fact, this outer green zone 7
also has gaps in it, because it's going to be in 8
pieces.
9 And so, all these gaps contribute to 10 bypass flow. It has a benefit, though. These bypass 11 flow paths also improve heat transfer during loss-of-12 forced-cooling accidents. So, while our predominant 13 heat transfer method is by radiation heat transfer 14 from the fuel assemblies to the reflector, we also get 15 a natural circulation from the upper plenum down 16 through the outer reflector bypass flow paths, and 17 then, into the fuel and back up, into the fuel here 18 and back up.
19 So, the bypass flow has a negative effect 20 during normal operation because it reduces the amount 21 of flow through the core, but it has a positive effect 22 during loss-of-forced-cooling accidents because it 23 aids in the natural circulation of helium through the 24 core.
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52 Any questions on this?
1 MEMBER KIRCHNER: Can you give us, John --
2 this is Walt Kirchner -- just some estimate of steady 3
state? What's the centerline temperature in your peak 4
rod bundle, roughly? You know, compared to LWR, the 5
cooling is not as efficient as a forced-water 6
circulation system. So, I'm imagining that your delta 7
T that's building up with the UO2 pellet on the order 8
of diameter of an LWR fuel rod would have a centerline 9
temperature that's running significantly higher. Is 10 your power density so low --
11 MR. BOLIN: Let's go back --
12 MEMBER KIRCHNER: -- versus your surface 13 area, that the centerline uranium oxide temperatures 14 are low?
15 MR. BOLIN: Well, let's go back to the 16 chart I had.
17 So, yes, you can see here that the fuel 18 rod linear power is about -- it's not quite 10 times 19 lower, but it's close, almost 10 times lower linear 20 power than an LWR. So, the delta T in the fuel is, 21 correspondingly, 10 times lower. And so, our peak 22 fuel temperatures are running around, I think around 23 1200 degrees C.
24 MEMBER KIRCHNER: Okay. It doesn't quite 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
53 scale like that. I'm sorry.
1 MR. BOLIN: Well, it doesn't quite --
2 MEMBER KIRCHNER: You've got a helium 3
coolant and --
4 MR. BOLIN: The helium -- the cladding 5
temperature is a lot higher.
6 MEMBER KIRCHNER: Yes.
7 MR. BOLIN: Yes, the cladding temperature 8
is -- obviously, the cladding temperature is higher 9
than the helium that's cooling it. But, if the 10 cladding temperature is around 800 degrees C, we have, 11 like I said, a much lower fuel delta T, and it, 12 correspondingly, lowers our peak fuel temperature.
13 MEMBER KIRCHNER: It would be useful to 14 have simple graphs at least for steady-state full-15 power conditions what your centerline UO2 16 temperature; what your cladding temperature is; what 17 the coolant is.
18 MR. BOLIN: Yes, I have that.
19 MEMBER KIRCHNER: And then, for a loss of 20
-- I think Dennis asked earlier -- for a loss of 21 forced circulation, where this was a real issue for 22 some of the fast reactors, gas-cooled fast reactor 23 designs, what kind of temperature excursion you would 24 see in a loss-of-flow condition?
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54 MR. BOLIN: Yes, I think that is a subject 1
that is best addressed later, I mean in a subsequent 2
discussion maybe. I actually do have some slides on 3
that, but it is for a different meeting. So, I didn't 4
put it in this package.
5 So, for accident conditions, we are --
6 MEMBER KIRCHNER: These comparisons that 7
you're showing aren't very useful, actually.
8 MR. BOLIN: Well, yes, I mean, they're --
9 MEMBER KIRCHNER: Because I think you get 10 my point. I mean, it's a rather complicated thermal 11 hydraulic set of conditions that doesn't extrapolate 12 well from gas to water.
13 MR. BOLIN: Correct.
14 MEMBER KIRCHNER: It depends on how much 15 surface area you have; what your ilium flow velocity 16 is --
17 MR. BOLIN: So, the surface --
18 MEMBER KIRCHNER: -- a whole number of 19 things. So, when you just do this apples-and-oranges 20 comparison, it's not terribly useful. It would be 21 much more for your benefit in making your case to show 22 what the actual operating conditions are for the fuel 23 and the cladding, and how the system responds in a 24 loss-of-flow condition.
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55 MEMBER MARCH-LEUBA: Yes, this is Jose.
1 Eventually, you're going to have to choose 2
your licensing basis events and run all the Chapter 15 3
analyses and figure out what your temperatures are 4
everywhere. But, before I invest money in the design, 5
I assume you have run some preliminary calculations 6
for what you consider to be the limiting event.
7 MR. BOLIN: Correct, we have.
8 MEMBER MARCH-LEUBA: And I would expect 9
that likely to be loss of pressure and when you lose 10 your gas.
11 MR. BOLIN: Correct.
12 MEMBER MARCH-LEUBA: And then, you started 13 cooling off by radiation. What temperatures you reach 14
-- are you going to be okay?
15 MR. BOLIN: Correct, we are. We've 16 designed it for that accident.
17 And as far as the surface area is 18 concerned, remember, our fuel rods have the same 19 geometry as the light water reactor fuel rods. So, 20 our fuel rod surface area is the same as the light 21 water reactor.
22 But that's about, you know, in the future, 23 you will need to present both normal operation and 24 accident condition fuel performance.
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56 So, we have -- and I know the NRC project 1
management is also going to go over this -- but we 2
have made a lot of progress in pre-application 3
licensing with the NRC. We have prepared a Regulatory 4
Engagement Plan, you know, that outlines our licensing 5
strategy for this conceptual period.
6 Obviously, the subject of this discussion 7
is the principal design criteria. We submitted that; 8
got some requests for additional information, and we 9
responded to those and revised the Topical Report.
10 And that's the subject of this meeting, is the NRC's 11 Safety Evaluation on that Topical Report.
12 We have also submitted a QA program 13 description Topical Report. We got some feedback 14 early on, and then, made changes to that document.
15 And that's undergoing review now.
16 And we've also submitted a
Fuel 17 Qualification Plan Technical Report. It's still a 18 fairly early document. It's like a white paper. It's 19 not a Topical Report.
20 And then, we have other documents that are 21 planned: a Source Term Methodology, LBE selection, 22 PRA, and safety classification.
23 CHAIR BIER: A quick question here.
24 You're expecting that the safety classification 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
57 decisions will be made based on the PRA or they are 1
two separate processes?
2 MR. BOLIN: Well, we will not have a 3
complete PRA at the end of this conceptual design 4
period. But the safety classification will be 5
informed by the preliminary risk assessment work that 6
is being done by, actually, being done by Vanderbilt 7
University -- and also, historical PRA work that we've 8
done on gas-cooled reactors in the past. So, it will 9
be risk-informed safety classification. And the LBE 10 selection will also be risk-informed, but not based on 11 a complete PRA.
12 CHAIR BIER: Okay. Thanks.
13 MEMBER KIRCHNER: John, this is Walt 14 Kirchner.
15 MR. BOLIN: Yes?
16 MEMBER KIRCHNER: I started to digress or 17 regress. Your basic reactivity control in terms of 18 accident conditions is based on leakage? In other 19 words, the diameter of your core and power level were 20 chosen or are pretty much indirectly determined 21 depending on leakage?
22 You've gone away from reflector control to 23 control rod controls. So, in an offsite condition, 24 what's the primary shutdown mechanism for this? Is it 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
58 Doppler and leakage? Could you give us just a feeling 1
for your core design philosophy in terms of reactivity 2
control?
3 MR. BOLIN: Well, while Doppler is a 4
factor that can lessen the reactivity control 5
requirements, we are still primarily relying on 6
control rods and shutdown rods. So, we have control 7
rods that will be, you know, partially inserted into 8
the core for power adjustment, but we also have 9
shutdown rods that will be fully removed from the core 10 that will be used for shutdown, out-of-steam shutdown 11 rods.
12 So, leakage and Doppler and reactivity 13 coefficients, those will all play a factor, but we're 14 still primarily relying on control rods and shutdown 15 rods.
16 MEMBER KIRCHNER: I'm thinking to offsite 17 conditions.
18 MR. BOLIN: Yes.
19 MEMBER KIRCHNER: So, our fast sodium 20 reactors, typically, are relying on leakage -- that 21 determines the maximum size of the core -- and 22 expansion.
23 MR. BOLIN: Yes, we're not relying on 24 expansion or --
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59 MEMBER KIRCHNER: Right.
1 MR. BOLIN: Like I said, leakage is not --
2 I mean, it's there, but we're not relying on leakage.
3 MEMBER KIRCHNER: Are your reflectors, 4
your radial reflectors, positive or neutral? Or 5
negative?
6 MR. BOLIN: I don't know the answer to 7
that.
8 MEMBER KIRCHNER: It's just something to 9
think about.
10 MR. BOLIN: Okay.
11 MEMBER KIRCHNER: We'll ask in a future 12 engagement.
13 MR. BOLIN: Yes.
14 MEMBER KIRCHNER: Thank you.
15 MEMBER PETTI: So, John, then, what 16 determines the size? Why is it 110 megawatts? Was 17 there some accident that limited, you know --
18 MR. BOLIN: Definitely. It definitely was 19 the depressurized loss-of-forced-cooling accident that 20 determined the size. And, in fact, we originally were 21 looking at 50 megawatts electric and 120 megawatts 22 thermal. And we ended up reducing the power to 100 23 megawatts thermal for the depressurized loss-of-24 forced-cooling transient.
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60 MEMBER PETTI: And is it a vessel 1
temperature or a peak silicon carbide temperature?
2 MR. BOLIN: It was peak silicon carbide 3
temperature.
4 MEMBER MARCH-LEUBA: Am I to assume from 5
what you said that your safety margin is 20 percent?
6 I mean, you cannot handle 50 megawatts electric, but 7
you can handle 40? I mean, that's very limited for an 8
advanced reactor.
9 Yes, I'm just putting it in the record.
10 You don't have to answer it. But we're used to seeing 11 reactors that you can shoot them with a shotgun and 12 nothing happens to it.
13 MR. BOLIN: Yes.
14 MEMBER MARCH-LEUBA: Are you telling me 15 that you want to go out to 20 percent power operate 16 and make it?
17 All right. Don't answer. I'll put the 18 bad things on the record, but you don't have to --
19 MR. BOLIN: I would say that the analysis 20 has improved during the conceptual design period, but 21 we have decided to stick with the 100 megawatts rather 22 than try to minimize our margin.
23 And silicon carbide has different --
24 MEMBER KIRCHNER: Just to calibrate us, 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
61 John, what do you use as a benchmark for your silicon 1
carbide structure as a thermal, like a thermal limit, 2
to determine, you know, it remains intact in a loss-3 of-depressurization and loss-of-forced-circulation 4
event?
5 Because that's, typically, you know, if 6
you go back to the HTGR business that you all were 7
involved in, you know, sizing of the core was kind of 8
an inverse calculation of temperature of the vessel 9
and temperature of the TRISO particle fuel.
10 What limits here? If it's silicon 11 carbide, can you calibrate us? What temperature is 12 that?
13 MR. BOLIN: The temperature limit we have 14 been using is 1800 degrees C. So, we want our silicon 15 carbide to be below 1800 degrees C.
16 MEMBER MARCH-LEUBA: And that's because of 17 the welding on the top of the rod?
18 MR. BOLIN: No. No.
19 MEMBER MARCH-LEUBA: So, I thought that 20 was the number that we were given earlier.
21 MR. BOLIN: The cladding itself. No, it 22 doesn't, it doesn't --
23 MEMBER PETTI: Decomposition? Because I 24 always thought decomposition was a little higher than 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
62
-- around 2000.
1 MR. BOLIN: Yes, decomposition is higher 2
than that.
3 MEMBER PETTI: Yes.
4 MR. BOLIN: It's higher than that.
5 MEMBER PETTI: So, it's just a composite 6
7 MR. BOLIN: It is degradation. I mean, 8
there's degradation of the cladding at 1800 C. And 9
so, we want to stay below that point. And we are.
10 MEMBER REMPE: I am confused because, 11 earlier, Aaron said that the peak temperatures for the 12 end caps for air are -- or I guess for helium -- was 13 1900 C?
14 MR. BOLIN: Correct.
15 MEMBER REMPE: So, what are the -- aren't 16 the end caps made of silicon carbide, too?
17 MR. BOLIN: Yes. Yes.
18 MEMBER REMPE: Was it 1800 or 1900?
19 MR. BOLIN: Eighteen hundred was just the 20 number we were using as a design --
21 MEMBER REMPE: So, you have margin, is 22 what you're saying?
23 MR. BOLIN: Yes. Yes.
24 MEMBER REMPE: Unless there's not any 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
63 steam in there.
1 MR. BOLIN: Right.
2 DR. SCHULTZ: John, this is Steve Schultz.
3 Just thinking about some operational considerations.
4 You mentioned the fuel cycle approach was going to be 5
such that the fuel assemblies would be in reactor for 6
a fairly extended period of operation. Any concerns 7
about the fuel assembly dimensional stability over 8
those long periods of time to high burnups as well?
9 In other words, are you going to have any problems 10 after many, many years moving assemblies when you do 11 your fuel management at infrequent intervals?
12 MR. BOLIN: Well, we will be moving fuel 13 assemblies every -- so, when we refuel every 15 years, 14 all the fuel assemblies get moved.
15 DR. SCHULTZ: We hope.
16 MR. BOLIN: Well, yes. And like I said, 17 15 years, we will have design gaps around the fuel 18 assemblies. And like I said, silicon carbide is very 19 rigid material. So, no deformation is --
20 DR. SCHULTZ: So, the accommodation is 21 there in the design --
22 MR. BOLIN: There is some --
23 DR. SCHULTZ: -- for that type of an 24 approach?
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64 MR. BOLIN: Correct. Correct.
1 MEMBER PETTI: How many dpa's, John, 2
about?
3 MR. BOLIN: Well, on average, it is 77.
4 MEMBER PETTI: And peak?
5 MR. BOLIN: No. And peak is about 100.
6 So, pretty aggressive. But it's been done before.
7 MEMBER PETTI: Yes. No, I'm not too 8
worried about it. There's data that shows that it's 9
okay. I'm more worried about how you qualify a 15-10 year fuel cycle in ATR, which is like super-11 accelerated for light water reactors. It's kind of 12 off the charts for this fuel.
13 MR. BOLIN: It's a subject of a 14 presentation I'm giving on Thursday.
15 MEMBER PETTI: Oh, good. I'm glad you've 16 got the answer.
17 DR. SCHULTZ: I am, too. That sounds 18 good.
19 MR. BOLIN: It's at INL. So, if you're in 20 the neighborhood --
21 DR. SCHULTZ: Well, thank you, John.
22 MEMBER REMPE: Back in D.C., though, a 23 question. Since this is your last slide, what are you 24 guys going to do for fuel ultimate storage? Is it 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
65 going to be onsite? Are there any special concerns 1
about how you're going to store it onsite, and then, 2
ultimately, if we ever have a repository, transferring 3
it to the repository? Or at least --
4 MR. BOLIN: So --
5 MEMBER REMPE: Go ahead.
6 MR. BOLIN: So, we're still working on 7
that. We're going to store onsite within the reactor 8
building, similar to other gas-cooled reactor designs.
9 At least have a core's worth -- I think core and a 10 reload worth of storage onsite in a spent fuel storage 11 area. Because I don't want to say -- I don't know if 12 it's going to be a pool, a vault, or storage wells.
13 That's still being worked out.
14 Eventually, I think we'll use dry storage 15 on the outside of the reactor building, similar to 16 like what light water reactors are doing now. We may 17 be able to move fuel into dry storage casks fairly 18 early because of our low power density. So, that's 19 still being investigated.
20 I do have to show my last slide just to 21 acknowledge that this work was supported by the 22 Department of Energy, Office of Nuclear Energy, under 23 that contract for advanced reactor concepts.
24 So, let's see.
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66 CHAIR BIER: So, we are remarkably close 1
to being on time. You did a good job anticipating how 2
many questions and comments you would get, I guess.
3 MR. BOLIN: Yes.
4 CHAIR BIER: If there are no more 5
questions on this presentation, this is probably a 6
good time to take a break, as scheduled, and be back, 7
I guess, at maybe 3:30, instead of 3:25. Is that okay 8
with people? I'm sorry, 2:30. I'm looking at the 9
wrong computer and doing the conversion in my mind of 10 time zones. You're right.
11 We're ahead of schedule. So, let's take 12 a break until 2:30, and then resume. Is that 13 agreeable to everybody?
14 (Whereupon, the above-entitled matter went 15 off the record at 2:12 p.m. and resumed at 2:32 p.m.)
16 CHAIR BIER: All right. Sorry for the 17 brief delay. Are people ready to move forward?
18 John, I believe you are up after the 19 break, also, for the principal design criteria, is 20 that correct?
21 MR. BOLIN: Correct. Can you hear me?
22 CHAIR BIER: Yes.
23 MR. BOLIN: And is my audio better?
24 CHAIR BIER: Well, it sounds fine so far.
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67 Okay.
1 MR. BOLIN: Okay. Because I have donned 2
a set of headphones and microphones. Hopefully, 3
that's going to --
4 CHAIR BIER: That will probably help, yes.
5 MR. BOLIN: -- cut back on ambient noise.
6 CHAIR BIER: Thank you. We appreciate it.
7 And so far, you are not sharing your slides. I 8
believe you're planning to.
9 MR. BOLIN: No, I've not started sharing.
10 CHAIR BIER: Okay.
11 MR. BOLIN: Yes.
12 CHAIR BIER: That's fine.
13 MR. BOLIN: Yes. Let me pull up that 14 presentation.
15 CHAIR BIER: By the way, I thought your 16 slides were quite readable, which is nice. Sometimes 17 they're minuscule eye charts, but these were pretty 18 good. Thank you.
19 MR. BOLIN: There might have been a few 20 tests, eye tests, on there.
21 Okay. Let's see here. Okay. How is 22 that?
23 CHAIR BIER: It looks good. Thank you.
24 MR. BOLIN: Okay. This is an overview of 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
68 the Principal Design Criteria for the Fast Modular 1
Reactor. Okay. There.
2 So, we did use the NRC guidance in 3
adapting and developing our Principal Design Criteria.
4 The Reg Guide 1.232, as you're aware, developed PDCs 5
for non-light water reactors by modifying and 6
supplementing 10 CFR 50, Appendix A, General Design 7
Criteria. And they did that in three categories:
8 sodium-cooled fast reactor, modular high-temperature 9
gas-cooled reactor, and then, a design-neutral 10 advanced reactor design criteria.
11 So, we used the ARDC and the MHTGR-DC as 12 starting points, and then, in our Topical Report, we 13 modified the NRC rationale for adaptation of the GDC 14 to our application for the FMR-DCs.
15 So, I'll just quickly go over some of the 16 key things about the FMR-DCs, and I've organized it in 17 the major categories of the design criteria.
18 So, the first category is overall 19 requirements, FMR-DC 1 through 5.
20 So, FMR-DC 1 is the same as the GDC.
21 Likewise, FMR-DC 2 is also the same as 22 GDC.
23 FMR-DC 3 is the same as the ARDC.
24 FMR-DC 4 made a slight change to 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
69 MHTGR-DC. We changed missiles. We expanded it or 1
made it more specific to include missiles originating 2
both inside and outside the reactor helium pressure 3
boundary. And we did that to explicitly cover any 4
missiles generated by the turbine machinery.
5 And then FMR-DC 5 is the same as the GDC.
6 The next category is multiple barriers, 7
FMR-DC 10 through 19. The FMR-DC 10 on reactor 8
design, the fuel design, as we went over previously, 9
the fuel design using SiGA cladding, it functions 10 similarly to light water reactors in that, you know, 11 that there's kind of a classic cladding function.
12 And so, we chose to use the SAFDL 13 terminology, both here and in other FMR-DCs. So, 14 where we may have been using a MHTGR-DC, we, 15 basically, chose to use the SAFDL terminology instead 16 of the -- I don't know how you pronounce it -- SARRDL 17 terminology.
18 Okay. Then, FMR-DCs 11, 13 through 15, 17 19 through 18, those are the same as the ARDCs and MHTGR-20 DCs, but with minor terminology changes. So, they're, 21 essentially, the same as those.
22 FMR-DC 12, the suppression of reactor 23 power oscillations, the word "structures" was added to 24 address reflectors, but the word "coolant" was 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
70 deleted.
1 So, as mentioned, the helium coolant 2
itself is neutronically neutral or inert or 3
transparent. And so, helium density flow changes, in 4
and of itself, don't cause a reactivity change, but --
5 MEMBER MARCH-LEUBA: This is Jose.
6 But, conceivably, you could have something 7
like u-tube momentarily-type oscillation of the 8
coolant between the reflector and the core, for 9
example. I don't know if you're supposed to go into 10 that. And that will, even though the helium is (audio 11 interference), it changes the temperature and has some 12 Doppler feedback. I don't suspect, I mean, it's even 13 remotely a problem, but you should analyze it.
14 MEMBER KIRCHNER: John, before you answer, 15 let me add on.
16 So, you have a fast reactor here, and you 17 have a reflector in the middle of it. Now, that's an 18 adaptation from the HTGR world. That's to push the 19 power out and to allow your passive heat rejection 20 system to take care of decay heat and manage the 21 vessel wall temperature, and a number of other 22 factors.
23 But, for a fast reactor now, does that 24 reflector decouple one part of the reactor from 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
71 another? In other words, do you have a tightly-1 coupled neutronic design or is this loose now -- I 2
use that word loosely -- in terms of how the core 3
might behave with regard to power oscillations?
4 Because you're now running in a fast spectrum, not a 5
thermal spectrum.
6 MR. BOLIN: So, I think the coolant, in 7
and of itself, is not a source of power oscillation.
8 But the flow is a possible source of power 9
oscillation. So, I think that's why structures were 10 added to address whether flow could cause the 11 reflectors to move, and therefore, cause a reactivity-12 generated power oscillation.
13 Now, that flow through the reflectors can 14 affect both position and temperature. So, I think 15 we're covered by that.
16 Since it is a fast-spectrum reactor, it 17 should be fairly tightly-coupled, and these 18 reflectors, like I said, this is zirconium silicide.
19 So, it's a fairly heavy reflector. So, it doesn't 20 moderate like a lot of reflectors tend to do, you 21 know, graphite or water, or whatnot.
22 So, the coupling, I think the coupling is 23 tight, but we're still looking at whether there's 24 really any significant oscillation or reactivity 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
72 feedback from the reflectors.
Of
- course, 1
historically, Fort Saint Vrain had a power oscillation 2
issue with the fuel columns and their moving around.
3 So, I think that's why we don't want to totally ignore 4
the structures.
5 Okay? Then, the last DC in this category 6
is containment design. And we are using the same as 7
the SFR-DC because the FMR uses, like we discussed 8
earlier, a low-leakage pressure-retaining containment.
9 So, more in line with the SFR-DC and, certainly, not 10 the vented confinement of the MHTGR.
11 The reactivity control is FMR-DC 20 12 through 29.
13 FMR-DC 20 through 24 are the same as the 14 GDCs.
15 FMR-DC 25 is the same as ARDC with minor 16 terminology changes.
17 And then, FMR-DC 26, just like the ARDC 18 and MHTGR-DC, it combines GDC 26 and GDC 27.
19 FMR-DC 28 is the same as MHTGR-DC.
20 And FMR-DC 29 is the same as GDC.
21 Fluid systems, FMR-DC 30 through 46.
22 30 through 33 are the same as MHTGR-DC.
23 The FMR-DC 34, residual heat removal, it's 24 similar to MHTGR-DC, but we wanted to make sure 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
73 it covered both the active, non-safety-related and 1
passive safety-related systems available to remove 2
residual heat. Also, similar to MHTGR-DC, it 3
incorporates the requirements in GDC 35.
4 And then, FMR-DC 36 and 37 are the same as 5
MHTGR-DC.
6 38 through 41 are the same as the ARDC.
7 DC-42 is the same as GDC.
8 DC-43, 45, and 46 are the same as ARDC.
9 And FMR-DC 44 is the same as MHTGR-DC.
10 So, all of these PDC selections are driven 11 by the design choices that we've made in the design.
12 And I believe this is the last slide of 13 DCs. It is reactor containment.
14 So, 50 through 53 are the same as ARDC.
15 54 is the same as SFR-DC.
16 And 55 through 57, they're the same as 17 ARDC, but with minor terminology changes.
18 And then, the next category is fuel and 19 radioactivity control.
20 60, 62, and 63 are the same as GDC.
21 And 61 is the same as ARDC.
22 And the same acknowledgment as the 23 previous presentation. It's supported by the U.S.
24 DOE, Office of Nuclear Energy, under that contract.
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
74 So, that was that.
1 CHAIR BIER: So, this is Vicki Bier. I 2
have a couple of very general, high-level questions.
3 One, could you discuss briefly, of the 4
modifications you described, which ones were kind of 5
the most crucial for safety versus just matching the 6
terminology to what's in your design?
7 MR. BOLIN: Well, let's see here.
8 MEMBER MARCH-LEUBA: How about DC 4 with 9
the missiles?
10 MR. BOLIN: Yes, that probably is the most 11 unique challenge from the FMR, is the missiles.
12 Because that, obviously, adds -- I mean, not that it 13 wouldn't have been considered, anyway, but it 14 certainly adds a design focus. I mean, not that we 15 would have ignored it, but, yes.
16 CHAIR BIER: Sure. One other, again, 17 high-level question. I know that the Reg Guide says 18 that it is possible for the applicants to identify 19 entirely new PDCs for unique features of the design 20 that are not adequately covered by the kind of 21 templates in the Reg Guide. And did you find any 22 situations where you were at least pondering that or 23 thought it might be worthwhile? Or do you think 24 you're close enough to the samples in the Reg Guide, 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
75 that you were able to cover what you needed there?
1 MR. BOLIN: I think, because the staff 2
covered the two different, very different, advanced 3
reactors -- the sodium fast reactor and the MHTGR --
4 that I don't think there were -- we did not identify 5
any gaps in design criteria that we -- we did not 6
identify any gaps.
7 CHAIR BIER: Okay. So, in other words, 8
the reason why what you have looks a lot like the 9
samples in the Reg Guide is really just because the 10 Reg Guide is pretty thorough and comprehensive, not 11 because you were just going through kind of a checkbox 12 process of "pick one from each column" kind of thing?
13 MR. BOLIN: Correct. Correct.
14 CHAIR BIER: Okay.
15 MR. BOLIN: I mean, I think the Reg Guide 16 was extremely useful in this process.
17 MEMBER REMPE: Well, thinking about the 18 historical approaches that GA has developed, where you 19 start with the critical safety functions, are you 20 doing that or applying that approach with this design?
21 Is it just, you know, control radionuclide release --
22 MR. BOLIN: It's the same critical safety 23 functions, correct.
24 MEMBER REMPE: So, there's nothing 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
76 unique or different, then, when you think about this 1
design? I mean, sometimes chemical reactions comes in 2
with higher priority, but you just didn't see anything 3
else?
4 MR. BOLIN: No. I mean, obviously, we 5
still have some graphite. So, we do have graphite 6
concerns. But we don't have -- the graphite is not in 7
the high-temperature parts of the core.
8 We still have water ingress concerns, but 9
we don't have high-temperature, high-pressure steam.
10 So, a lot of our safety concerns from MHTGR are quite 11 a bit lessened.
12 MEMBER MARCH-LEUBA: This is Jose.
13 How about the very long cycle time, the 14 15-years recycle/reload, and the implications that you 15 may have on misalignment of fuel, clipping, phase-in, 16 moving, vibrations? And 15 is a long time before you 17 open and look inside to see what's going on.
18 MR. BOLIN: Yes.
19 MEMBER MARCH-LEUBA: I mean, does that 20 affect something?
21 MR. BOLIN: Certainly, we do expect to 22 have to shut down more frequently than every 15 years 23 for other maintenance and inspection reasons.
24 MEMBER MARCH-LEUBA: Yes, but do you 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
77 expect to open the core? You'll probably be fixing 1
some pump outside of the core plenum.
2 MR. BOLIN: Well, that hasn't been 3
decided, whether --
4 MEMBER MARCH-LEUBA: Yes. And if I 5
incorrectly wanted to --
6 MR. BOLIN: And, you know, for the first-7 of-a-kind prototype, it might be you might do some 8
fuel inspection.
9 MEMBER MARCH-LEUBA: I'm just trying to 10 think what is different. If I read correctly your 11 cartoons, the control rods are sitting outside, or the 12 shutdown rods for sure are sitting outside the vessel, 13 and they have to go through sealed?
14 MR. BOLIN: Well, the control rod drive 15 mechanism and connecting rod are out -- well, they're 16 not -- technically, that's still part of the vessel.
17 It's still part of the helium reactor pressure 18 boundary.
19 MEMBER MARCH-LEUBA: Then, enclosed? I'm 20 just wondering if there is some design criteria that 21 applies to those special configurations.
22 MR. BOLIN: Well --
23 MEMBER MARCH-LEUBA: It certainly feels --
24 let me put it this way; I'm the bad guy here -- it 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
78 simply feels that you took all the GDCs that were in 1
the design guide and went through to see if they 2
applied to you, instead of thinking about your design 3
and see what's missing. It's something very human to 4
do, and that's something we all do.
5 So, on the review, I'll be asking the 6
staff, when they're here, if they thought, what's 7
missing?
8 MR. BOLIN: Okay.
9 MEMBER MARCH-LEUBA: It's very easy, when 10 somebody gives you a paper, to correct the English, 11 but what's important is, what paragraph is missing in 12 that article? The same thing here.
13 MR. BOLIN: All right.
14 CHAIR BIER: For operational reasons, are 15 you anticipating that there would be periodic 16 shutdowns for reasons other than refueling, or that 17 it's just going to run flat-out and just adjust power 18 levels?
19 MR. BOLIN: I mean, it hasn't been worked 20 out specifically, but there is discussion of whether 21 we would want to shut down every five years for an 22 inspection, particularly, you know, power conversion 23 unit inspection and/or generator or control rod drive 24 motors, or a variety of things we might want 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
79 inspect every five years. At least, particularly, we 1
have to consider increased inspection frequency for a 2
first-of-a-kind plant. So, that's being discussed.
3 MEMBER PETTI: Wouldn't Section 11 4
require, like, the vessel to be inspected?
5 MR. BOLIN: Yes, vessel inspection is 6
another example.
7 MEMBER HALNON:
- Yes, and don't 8
underestimate the power of the insurance agency.
9 MR. BOLIN: Well, yes. We tend to ignore 10 that until the very end.
11 CHAIR BIER: Are there other questions or 12 comments for John, or any other points that John wants 13 to add, before we transition to the staff?
14 (No response.)
15 CHAIR BIER: I guess one other question 16 that I have, I noticed that there was a Rev 1 of the 17 PDCs, which I guess was in response to the RAIs from 18 the staff; that some things got adjusted? Again, are 19 there any there that are noteworthy enough that you 20 want to call them out or discuss the value of those 21 changes?
22 MR. BOLIN: Well, it's interesting that a 23 lot of the changes were -- we had actually prepared 24 our DCs based on a draft of the Reg Guide. And then, 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
80 the Reg Guide changed. And so, there were 1
inconsistencies between our DCs and the Reg Guide.
2 So, a lot of the corrections were just making those 3
corrections to the revised Reg Guide.
4 CHAIR BIER: Okay. If there are no 5
further questions and comments, then, I guess we can 6
transition to staff. And I'm not sure if the primary 7
presenter is Reed Anzalone or Samuel Cuadrado. Which 8
of --
9 MR. ANZALONE: It's going to be me.
10 CHAIR BIER: Okay. Thank you.
11 So, I guess, John, you can stop sharing 12 your slides, then. Thank you very much for the 13 presentation.
14 MR. BOLIN: Thank you.
15 MR. ANZALONE: Okay. Thanks. So, thank 16 you, everyone, for having us here today.
17 My name is Reed Anzalone. I'm a Senior 18 Nuclear Engineer in NRR's Division of Advanced 19 Reactors and Non-Power Production and Utilization 20 Facilities. I'm joined today by our Project Manager, 21 Sam Cuadrado, who is also with me in DANU.
22 So, I was the lead technical reviewer for 23 this effort, and I was assisted by Sheila Ray in our 24 Division of Engineering and External Hazards, who'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
81 here on the phone. She covered the electrical PDCs, 1
and Steve Jones, who is with me in DANU, covered the 2
containment PDCs, who wasn't able to make it. So, if 3
there are questions about those, I can address them.
4 Next slide, please.
5 So, quick agenda, and you'll see that a 6
lot of this should look very, very, very, very 7
familiar from the presentations that we just had from 8
General Atomics. I was laughing the whole time during 9
John's presentation because there is almost a one-to-10 one correspondence between the topics covered. So, I 11 may go quickly through some of these. And, of course, 12 if you have questions, feel free to interrupt.
13 MEMBER MARCH-LEUBA: Is that the 100 14 percent rule to coordinate with the other presenter?
15 MR. ANZALONE: No. In fact, we --
16 MEMBER MARCH-LEUBA:
In
- fact, the 17 desirable?
18 (Laughter.)
19 MR. ANZALONE: We only got the slides 20 yesterday. So, I was happy to see that they matched 21 very well.
22 So, Sam will be talking a little bit about 23 the pre-application engagement. Then, it will go back 24 to me, and I'll talk about the Topical Report 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
82 timeline. That's really just going to be for 1
reference, for anyone who might want to go back and 2
look at the correspondence that we had.
3 I'll touch a little bit on some of the 4
design features that we already talked about; talk a 5
little bit about the PDC guidance that's out there; 6
the PDC development approach that General Atomics 7
provided to us in their Topical Report, and then, I'll 8
go into the Fast Modular Reactor design criteria 9
themselves, including kind of highlighting the key 10 design choices and the effects that those had on the 11 PDCs. And hopefully, I can address the question that 12 you raised. And then, I'll just briefly touch on the 13 Safety Evaluation and conclusions.
14 So, Next slide. MR. CUADRADO DE JESUS:
15 Now, good afternoon. Sam Cuadrado.
16 So, this is a brief overview of the pre-17 application engagement with General Atomics. You saw 18 a similar slide when John Bolin was doing the 19 presentation.
20 We got the pre-application letter 21 engagement plan last year in March. Accordingly, 22 we're reviewing a couple of documents, which include 23 this one that we see, the Topical Report which is the 24 topic of this meeting.
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
83 We have a few qualifications in the 1
Topical Report. Basically, it's only that we are 2
providing feedback in the form of a white paper. And 3
last month, we received the Quality Assurance Program 4
Topical Report. That's currently going through the 5
review process.
6 We expect a few more documents, a few more 7
submittals this year, and a couple more next year.
8 This year, for the summer, we've got the mechanistic 9
source term. By the end of the year, we should be 10 getting the licensing basis event white paper. And 11 for the spring of next year, the safety approach on 12 the PRA and safety classification white papers.
13 So, back to Reed on this.
14 MR. ANZALONE: All right. Next slide.
15 So, just quickly on the review timeline, really, all 16 I wanted to highlight here was that we did ask a round 17 of RAIs and we got prompt responses, and then, 18 subsequently, General Atomics rev'd the Topical Report 19 to incorporate those responses. And then we issued 20 the Draft SA in March.
21 Next slide. Well, go ahead.
22 MEMBER BALLINGER: I have a question. I 23 see that the Fuel Qualification Plan Technical Report, 24 it says, "under review." Do we know when we're going 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
84 to get that?
1 MR. CUADRADO DE JESUS: Yes. That is for 2
a white paper there. It was just some feedback. So, 3
they provided some "asks," some questions for us to 4
provide them feedback. I placed that information for 5
you guys to get access to it. But we plan to provide 6
feedback to them by November of this year.
7 MEMBER BALLINGER: So, we have access to 8
this?
9 MR. CUADRADO DE JESUS: You have access to 10 the request, yes, to the Technical Report.
11 MS. DE MESSIERES: This is Candace de 12 Messieres.
13 I just wanted to clarify that this is a 14 Technical Report, not a Topical Report.
15 MR. CUADRADO DE JESUS: Yes, yes, yes.
16 MS. DE MESSIERES: So, I just wanted to 17 make sure that that was clear.
18 MR. CUADRADO DE JESUS: Yes, but you can 19 see the Technical Report and the questions that they 20 want us to answer, to provide feedback. It's in 21 SharePoint for you guys.
22 MEMBER MARCH-LEUBA: But it has been 23 provided on the docket? I mean, we can see it?
24 MR. CUADRADO DE JESUS: Yes, it's on 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
85 docket.
1 MEMBER MARCH-LEUBA: But, typically, 2
Technical Reports are part of the SAR.
3 MEMBER HALNON: Yes, we, typically, don't 4
-- we separate white papers from Technical Reports.
5 This has both. Is it a white paper or is it an 6
actually approved -- since they don't have a QA 7
program yet, it can't be a Technical Report that they 8
would reference.
9 MR. ANZALONE: No, it's a white paper.
10 MEMBER HALNON: Okay.
11 MR. ANZALONE: It's just called a 12 Technical Report.
13 MEMBER HALNON: Yes, that's the title.
14 (Laughter.)
15 MR. CUADRADO DE JESUS: Yes, but it's on 16 the docket, so you guys can see it.
17 MR. ANZALONE: So, it's not going to be, 18 you know -- it doesn't get that stamp of finality 19 that's --
20 MEMBER HALNON: It won't be referenced out 21 of the SAR --
22 MR. ANZALONE: No.
23 MEMBER HALNON: -- whenever that comes.
24 CHAIR BIER: Just for completeness or 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
86 clarification, Weidong, if I understand correctly, 1
this is not in the SharePoint for this meeting, but 2
it's available through NRC, is that correct?
3 MR. WANG: Correct. That is, the staff 4
has created a SharePoint, but I think that Sam has --
5 MR.
CUADRADO DE JESUS:
- Yes, in 6
SharePoint, there's a folder related to General 7
Atomics.
8 MEMBER HALNON: So, on your next slide, 9
Rev 2 was transmitted. They make it Rev 1.
10 MR. ANZALONE: I think it should be Rev 1.
11 Sorry.
12 Okay. Next slide. So, just talking a 13 little bit about the design features, I know we've 14 just, literally, had a presentation from them. I just 15 wanted to kind of go through the things that we 16 thought were particularly noteworthy in our review.
17 So, one is, obviously, the core 18 arrangement, which is different from the other gas-19 cooled reactors that we've seen recently at the NRC, 20 which, you know, they're using, essentially, what 21 looks like an LWR core with the fuel rods and UO2 22 cladding, the silicon carbide. But the arrangement is 23 a little bit more like a fast reactor core with a 24 tight space in between the rods and triangular 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
87 tension, the hexagonal assemblies.
1 And when you compare that -- so, I was 2
thinking about things in terms of the basis for what's 3
in Reg Guide 1.232. So, that's the MHTGR, which was 4
a prismatic block gas-cooled reactor using TRISO fuel.
5 So, obviously, pretty different there.
6 The other thing that's big that you can 7
see on this slide is the gas turbine. I think that's 8
been covered pretty well.
9 And then, the MHTGR used steam generators 10 rather than having the power conversion system 11 directly on the primary circuit.
12 Next slide. The thing that I'll highlight 13 here, you see the containment. John talked a little 14 bit about that. So, there is an actual containment 15 building versus like a functional containment or 16 confinement approach.
17 And the other thing is the RVCS cooling 18 system, which is a little different from the passive 19 cooling system that was in the MHTGR design.
20 Yes, that's everything I wanted to cover 21 on this slide.
22 So, then, the key design features, and on 23 a future slide, I'll kind of talk about how these feed 24 into the PDCs. So, it just sort of sums up everything 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
88 that I covered in the last couple of slides. I don't 1
think I need to really talk any more about this.
2 Next slide. MEMBER KIRCHNER: Can you 3
just pull your microphone up? Maybe we can hear it 4
better. Yes, great.
5 MR. ANZALONE: All right. Sorry. Thanks.
6 Oh, that is much louder.
7 MEMBER MARCH-LEUBA: There's people on the 8
other side of the phone line that would love to see 9
how this works.
10 MR. ANZALONE: Yes. Thanks. That's much 11 better.
12 So, just a little bit of what we used for 13 guidance in evaluating PDCs and you know the kind of 14 conclusions that we're trying to reach. So, both of 15 these quotes are from Part 50, Appendix A.
16 And that first one, the first statement 17 there is kind of the conclusion that we're trying to 18 reach: that the Principal Design Criteria established 19 the necessary design, fabrication, construction, 20 testing, and performance requirements.
21 And then, the second statement there talks 22 about this is the guidance that Part 50, Appendix A, 23 gives in establishing PDCs. So, the GDCs in Appendix 24 A aren't directly applicable to non-light water 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
89 reactors, but they are considered to be guidance in 1
establishing PDCs for non-light water reactors.
2 Next slide. And then, also, we have Reg 3
Guide 1.232, which we talked about a little bit. That 4
was issued in April of 2018. I think most of the 5
members were on the Committee when it was issued.
6 And it documents three sets of acceptable 7
PDCs. So, there's the advanced reactor DCs, which are 8
supposed to be generic and technology-inclusive, and 9
there's an asterisk there because it's technology-10 inclusive for certain technologies that we had in mind 11 when we were writing them. I don't think that you 12 could make one that is, you know, wholly generic that 13 would be of value really.
14 Then, there is the sodium-cooled fast 15 reactor DC, which really, I think, were made with the 16 PRISM reactor in mind, and the MHTGR-DCs, which were 17 made with the MHTGR which is a TRISO-fueled, helium-18 cooled, as I mentioned, prismatic block, graphite-19 moderated, high-temperature gas reactor.
20 So, that slide --
21 CHAIR BIER: Before you move on, when 22 these design criteria were developed, did you envision 23 in advance that people might be mixing and matching, 24 based on what fits their circumstance?
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90 MR. ANZALONE: Yes. And actually, there's 1
one -- one of the points coming up, especially PDC 16 2
talks about containment design criteria. That 3
explicitly in it says, "We envisioned that people 4
would pick the one that best suits their design here."
5 So, I think that was, clearly, a consideration.
6 And the thing that I want to kind of point 7
out -- and this gets a little bit at Jose's question 8
-- the FMR kind of neatly straddles all of these 9
categories. It falls kind of in between all of them.
10 So, I don't think that there's any real aspect of the 11 design that is so exotic that it wouldn't be well-12 encompassed by these design criteria.
13 And then, that is something that we 14 thought about, as we were going through and doing the 15 review, is, you know, are these adequate? And the 16 answer that we keep coming back to was, yes, it looks 17 like this covers what it needs to.
18 MEMBER MARCH-LEUBA: And an impracticality 19
-- if that's a word -- when we do the full Chapter 15-20 type analysis and Chapter 19, it will mean something.
21 It will pop up there.
22 MR. ANZALONE: Mm-hmm. And one thing that 23 I think is interesting about this particular review is 24 that they came to us with these PDCs very, very early 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
91 in the project. I mean, this was the second thing 1
that was submitted after the Regulatory Engagement 2
Plan. So, really, it's the PDCs came, and then, the 3
design -- I mean, it got to a certain level of 4
maturity to be able to establish what the PDCs ought 5
to be, but, ultimately, they will have to design the 6
reactor to meet these PDCs.
7 MEMBER HALNON: But is it, I mean, written 8
generically enough to create a fourth category in 9
regards to the Reg Guide?
10 (Laughter.)
11 MEMBER HALNON: I mean, when I went 12 through it, it seemed like there were pretty generic, 13 directly written to advanced reactors of this type.
14 MR. ANZALONE: So, I don't know that 15 that's really necessarily worth doing. Because I 16 think there's this vision that people would kind of 17 mix and match. The vast majority -- and I have a 18 summary slide as a backup slide -- the vast majority 19 are just straight from the ARDC with minor 20 modifications here and there.
21 MEMBER HALNON: Okay.
22 MR. ANZALONE: So, I think it's, you know, 23 within the envelope of stuff that we would expect 24 people to do with this Reg Guide.
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92 CHAIR BIER: And presumably, any future 1
designs may also have their own unique tweaks and not 2
fit exactly with --
3 MEMBER HALNON: But when I got done 4
reading it, one of the things that just popped into my 5
mind, it just felt like the fourth category was just 6
written, but I understand what you're saying. They're 7
close enough to all these other things.
8 MR. ANZALONE: Yes.
9 MEMBER HALNON: There's nothing really 10 unique or brand-new in there that would warrant a 11 special --
12 MEMBER MARCH-LEUBA: If anything worries 13 me along this design, it's the high temperature. But 14 they'll eventually know how to do it.
15 MR. ANZALONE: All right. Next slide.
16 And General Atomics covered this in their 17 presentation, but the concept that was conveyed to us 18 in the Topical Report was that they would start with 19 the Advanced Reactor Design Criteria. Then, if that 20 wasn't fully applicable, they would go look at the 21 other ones for direct adoption, and then, take the one 22 that was the most applicable and adapt or refine it to 23 match with the design.
24 Okay. Next slide. So, then, talking 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
93 about the key design feature effects on the Principal 1
Design Criteria. So, the fuel and the core really 2
kind of lead to the use of SAFDLs rather than SARRDLs, 3
which John mentioned in his presentation. And that 4
also, I think, goes along with the Containment 5
Principal Design Criteria that they ended up using and 6
the containment design. We generally kind of think of 7
those as going together. SAFDLs go with functional 8
containment; SARRDLs go with leak-tight containment or 9
controlled leakage.
10 The neutron spectrum fast, I think the big 11 takeaway there that we wanted to make sure was 12 included was to consider the effect of structures on 13 reactivity feedback, which, otherwise, I think the 14 ARDC includes this, but the GDC, if that were to be 15 adopted directly, does not.
16 And actually, that was an RAI that we 17 asked. Because, originally, what was in the Topical 18 Report didn't include structures in that PDC, and we 19 wanted to make sure that that was in there.
20 The helium coolant, so for that, the big 21 thing there was -- they're all out of order on my 22 paper here -- that affects a whole bunch of the 23 Principal Design Criteria. The big effect is to 24 remove considerations related to coolant inventory 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
94 control, and that's consistent with the modular high-1 temperature gas reactor design criteria. And that 2
means there's no PDC 35, which relates to emergency 3
core cooling systems. And the emphasis is placed on 4
the residual heat removal systems.
5 And then there's also they decided to 6
change, to be consistent, the reactor helium pressure 7
boundary, instead of reactor coolant pressure boundary 8
or reactor coolant boundary in the PDCs.
9 There wasn't any particular effect for the 10 gas turbine on the primary coolant. DC 4 has the 11 consideration of missiles generated from either inside 12 or outside the containment. That was actually 13 included originally in the MHTGR-DC 4. So, I wanted 14 to note that there.
15 The residual heat removal -- and John 16 mentioned this on his slides -- so, they adopted the 17 MHTGR passive residual heat removal PDCs, but, then, 18 they adapted them to remove passive, so that it would 19 encompass both their passive system and the active 20 non-safety-related system.
21 And that seemed appropriate to us to do.
22 You know, it just made it broader, so that it covers 23 a wider scope of systems; whereas, the MHTGR-DC really 24 just, specifically, covers the passive residual heat 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
95 removal systems.
1 And then, for containment, I already 2
mentioned that they got the leak-tight containment 3
building, and they adopted the standard containment 4
PDCs.
5 Next slide. So, I've, basically, already 6
covered most of these in what I just said, but we can 7
quickly go through this.
8 So, I listed out all of the design 9
criteria, and then, highlighted ones that I think are 10 worth mentioning, either because they had to make a 11 particular choice about where they went with it or 12 they've modified it in an interesting way.
13 So, I just finished talking about PDC 4.
14 It's noteworthy because they wanted to include 15 missiles generated inside the reactor helium pressure 16 boundary.
17 Next slide. So, 10, we've got the --
18 MEMBER BROWN: Can I ask you a question 19 about the missiles?
20 MR. ANZALONE: Sure.
21 MEMBER BROWN: They show in their prior 22 generation a gas turbine-driven, you know, the heated 23 helium-driven TGs. Is that a very high speed? Is 24 that a very, very high-speed? I mean, all plants have 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
96 a missile issue relative to their turbine generator 1
sets you have to consider. So, this looked like a 2
high-speed one, which would make it more critical in 3
terms of covering it. Is that the reason for the 4
emphasis here?
5 MR. ANZALONE: No. I think the reason is 6
just that, like, it is noteworthy that they have a 7
power conversion system that is inside containment.
8 MEMBER BROWN: Oh, okay. All right.
9 MR. ANZALONE: And it's part of the 10 reactor coolant boundary.
11 MEMBER KIRCHNER: It's part of the primary 12 cooling boundary.
13 MEMBER BROWN: Yes. Okay. So, that part 14 I missed. I missed that it was inside the coolant, 15 the primary boundary.
16 MR. ANZALONE: Yes.
17 MEMBER BROWN: My brain fried on it.
18 (Laughter.)
19 MEMBER BROWN: Thank you.
20 MR. ANZALONE: Go ahead.
21 So, I already mentioned the use of SAFDLs 22 rather than SARRDLs. I will also mention that 23 Criterion 10 uses, talks about heat removal, rather 24 than coolant. So, I think the effect that you were 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
97 talking about of, you know, potentially, oscillatory 1
coolant behavior -- I think that talking about heat 2
removal is appropriate there.
3 The same thing with we talked about power 4
oscillations. Coolant isn't mentioned there, but it 5
talks about the core. And so, I think that's 6
appropriately considered, enveloped by that design 7
criterion. And it does have structures in there in 8
talking about power oscillations.
9 So, if there's an effect of the 10 reflectors, that's covered under that design 11 criterion. Whether that effect is caused by the 12 behavior of coolant affecting the structures or it's 13 something in the inherent behavior of the structures.
14 MEMBER MARCH-LEUBA: Yes, I sense some 15 real thinking that you worry about the structures 16 because of mechanical vibrations or displacement; 17 whereas, it could be a temperature oscillation. I 18 find it very unlikely that will happen, but --
19 MR. ANZALONE: Yes.
20 MEMBER MARCH-LEUBA: -- you have to 21 consider that it will happen.
22 MR. ANZALONE: No, but, I mean, actually, 23 I will say, part of the reason that we asked about 24 structures was, you know, knowing fast reactors 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
98 that they're, basically, a coupled system, we felt 1
like that was important to include. We didn't know at 2
the time when we asked that question that all of the 3
structures are going to be made out of silicon 4
carbide. And so, they don't really move very much 5
during power maneuver.
6 MEMBER MARCH-LEUBA: But you have those 7
unusual bowing effects.
It's an oscillation 8
configuration, a crucial thing.
9 MR. ANZALONE: But we think it's covered 10 by just making sure that structures are considered in 11 there.
12 MEMBER MARCH-LEUBA: Okay.
13 MR. ANZALONE: Thirteen, that was one 14 where they used the helium pressure boundary instead 15 of the reactor coolant boundary. And I know it's 16 instrumentation and control, but, really, the main 17 distinguishing feature between all the different DCs 18 was what the coolant system looked like.
19 Containment design. John already covered 20 that, I think in sufficient detail.
21 And electric power systems, they used the 22 MHTGR design criterion, but they modified it to go 23 with SAFDLs instead of SARRDLs. That's appropriate to 24 be consistent with the other DCs.
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99 MEMBER MARCH-LEUBA: Do we envision 1
safety-grade power? Or there is nothing that needs to 2
be driven?
3 MR. ANZALONE: I can't remember off the 4
top of my head. I'm going to phone a friend.
5 (Laughter.)
6 MR. ANZALONE: To Sheila, do you think you 7
can answer that question? Or John?
8 MR. BOLIN: John can answer it.
9 We don't see a need for Class 1E backup 10 electrical generation.
11 Does that --
12 MEMBER MARCH-LEUBA: Yes. But you usually 13 have a couple of batteries for the control room, 14 right?
15 MR. BOLIN: Correct. Correct. We'll have 16 17 MEMBER MARCH-LEUBA: Non-safety grade?
18 MR. BOLIN: Or, you know, there will be 19 containment isolation valves, other isolation valves 20 that might need to function. Whether they're 21 electrical, by battery, or some other means is still 22 to be determined.
23 MR. ANZALONE: Next slide. So, 26. I 24 want to highlight here we have had some challenges 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
100 with some applicants in PDC 26. I'm not going to go 1
into that in detail here. But here, they adopted the 2
PDC, the language in advanced reactor design criteria.
3 As is, with one modification, to be 4
consistent with the GDC, they included the effects of 5
xenon. We don't think that xenon is going to be 6
particularly important in a fast reactor, but there 7
wasn't any reason not to include it.
8 MR. BOLIN: And I'll second that. We have 9
recently found that also to be the case, that xenon is 10 really not of any -- it has no impact to speak of.
11 MR. ANZALONE: But it's included in the 12 design criterion.
13 MR. BOLIN: But it's there. It's there.
14 MR. ANZALONE: So, if somehow it's found 15 to have an effect, it's covered. More broad is 16 actually okay.
17 So, consistent with the all of the sets of 18 design criteria in the Reg Guide, got rid of PDC 27 19 and incorporated it into 26.
20 MEMBER MARCH-LEUBA: Now that I see, I'm 21 asking not PDC, but criteria limits. Is there an 22 issue with rod ejection here? We have 7 megapascal.
23 Will that be a licensing basis event?
24 MR. ANZALONE: I would think so.
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101 MEMBER MARCH-LEUBA: Okay.
1 MR. ANZALONE: We haven't looked at the --
2 we haven't gotten in the level of detail of 3
understanding the control rod design or the control 4
rod drive systems.
5 MEMBER MARCH-LEUBA: It's likely my lack 6
of familiarity with fast reactors, but even events on 7
fast reactors bothers me a lot.
8 MEMBER KIRCHNER: It's a high-pressure 9
envelope and the control rod mechanism is part of the 10 envelope, inside the envelope. So, it's the same as 11 a PWR when it comes to rod ejection.
12 MEMBER MARCH-LEUBA: Mm-hmm.
13 MR. ANZALONE: But that is a distinction 14 to sodium fast reactors which are not high-pressure.
15 For the reactivity limits, they went with 16 the modular high-temperature gas reactor design 17 criterion because it fit the best with the coolant 18 system design that they have. Again, that was the 19 biggest distinguishing feature between all of them.
20 Next slide. So, getting into fuel 21 systems, I think John already -- I think we covered 22 this. So, 33 and 35 were removed, and that's 23 consistent with the modular high-temperature gas 24 reactor design criteria, and like I said, reflects 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
102 focus away from coolant inventory control and towards 1
residual heat removal. And then, those residual heat 2
removal PDCs were adjusted to not specifically mention 3
passive systems.
4 Next slide. There wasn't anything, in 5
particular, that I wanted to highlight about these, 6
but I saw that, on John's slides, he mentioned that 7
containment heat removal, cleanups, and events were 8
things that they were considering. We think that 9
those are encompassed by these design criteria, the 10 way that they're written.
11 Next slide. So, here, 54 I think is 12 interesting because they referenced the sodium fast 13 reactor design criteria. They made a change to it to 14 remove reactor, to signify that there are more 15 structures inside. So, normally, it says, "reactor 16 containment," but they want to say, hey, we've got a 17 lot inside containment, aside from just the reactor.
18 The power conversion system is inside containment.
19 So, it just says, "containment," rather than "reactor 20 containment."
21 But, aside from that, the interesting 22 thing about SFR-DC 54 is that it talks about not 23 necessarily having to isolate systems that penetrate 24 containment where you wouldn't expect a release path.
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103 And that's key for them to be able to have the water 1
tanks for the RVCS outside of containment with lines 2
that go into containment. And because you wouldn't 3
expect a release pathway to go through those pipes, 4
that's acceptable. And that's consistent with the 5
SFR-DC.
6 MEMBER MARCH-LEUBA: And with 55, the 7
helium pressure boundary doesn't cross containment, 8
does it?
9 MR. ANZALONE: No, it does not.
10 MEMBER MARCH-LEUBA: I mean, there might 11 be some feedline.
12 MR. ANZALONE: Yes.
13 MEMBER MARCH-LEUBA: But that wouldn't 14 enter part of the pressure containment?
15 MR. BOLIN: I will correct. There is a 16 system that has helium in it that is connected to the 17 pressure boundary that does cross the containment 18 boundary, and that's the helium purification system.
19 MR. ANZALONE: Right.
20 MEMBER MARCH-LEUBA: Yes, but that would 21 be a small line.
22 MR. ANZALONE: Yes.
23 MR. BOLIN: It will be a small line, and 24 it certainly will have isolation valves on it.
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104 MR. ANZALONE: Well, and that's what the 1
design criterion says, that you need to do this when 2
you have reactor helium systems penetrating it.
3 MR. BOLIN: And we will do that.
4 MR. ANZALONE: And then, the next slide.
5 So, nothing particular with these. They all adopted 6
the advanced reactor design criterion as written.
7 Next slide. So, I just wanted to talk 8
briefly about the conclusions and the Safety 9
Evaluation. So, we think that they, appropriately, 10 considered the Reg Guide and developed a sufficient 11 set of PDCs that were appropriate for establishing the 12 requirements for the FRM design. And like I said, 13 they came early. So, these will be criteria that 14 we'll open to, as they continue interactions with us.
15 And what the SE says is that they 16 establish the necessary
- design, fabrication, 17 construction, et cetera, that 10 CRF 50, Appendix 50, 18 kind of establishes as the requirement for Principal 19 Design Criteria.
20 And then, I wanted to make note that the 21 Topical Report can be used by future applicants for 22 the FRM, but the way that we do Topical Reports, you 23 know, you have to justify the applicability of the 24 Topical Report when you come in and you use it. 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
105 so, there isn't a specific limitation and condition 1
that says it has to be like this reactor. But we 2
expect that, if somebody were to use this Topical 3
Report and reference it, they would have to justify 4
that if it was substantially different in any way, why 5
it was okay to use this design.
6 MEMBER MARCH-LEUBA: By "somebody," do you 7
mean like a different company?
8 MR. ANZALONE: Well, presumably -- so, 9
it's for the FRM design. So, I don't know -- some 10 other company bought that design from General Atomics 11 or if they spun off a subsidiary or --
12 MEMBER MARCH-LEUBA: Doesn't GA own the 13 intellectual property on the Topical Report? I mean, 14 nobody can use it without GA's permission. Well, it's 15 static.
16 MR. ANZALONE: And that's my last slide.
17 I have some backup slides that go over some of these 18 in more detail.
19 CHAIR BIER: So, again, a different 20 version of the same question that I asked John earlier 21 with regard to the RAIs. How many of those subsequent 22 changes to Rev 1 were because the Reg Guide itself 23 changed? How many were kind of minor editorial 24 improvements? And were there any that you thought 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
106 were really (audio interference)?
1 MR. ANZALONE: I don't think there were 2
any that specifically -- so, the one that I mentioned 3
earlier, which was including structures in the power 4
oscillations, that was an RAI, and that was, we think, 5
important to capture something that was missing.
6 I think most of the rest of them were, 7
hey, you said you're using this design criterion from 8
the Reg Guide, but the words that you're using don't 9
match up. And that could reflect what John said, that 10 they were using the draft version of the Reg Guide.
11 I think that covers pretty much all the RAIs between 12 those two.
13 CHAIR BIER: Are there any other questions 14 for Reed and Sam? Sorry. Yes, are there other 15 questions for Reed and Sam? Are there in the room or 16 online?
17 MEMBER KIRCHNER: I'd just make an 18 observation or two.
19 I mean, you asked both the applicant and 20 the staff -- my take, the major thing that's different 21 here is that, you know, this concept is straddling the 22 MHTGR and the fast, as has been pointed out, the fast 23 reactor PDCs and the MHTGR. So, the MHTGR, as we 24 know, is a functional containment approach -- SARRDLs, 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
107 if I got the acronym right, where this is SAFDLs and 1
a containment.
2 And I think the main issue that I see 3
coming is where the systems straddle the containment 4
boundary for this reactor. So, you've got quasi-5 passive systems, if I could call them that. The MHTGR 6
was meant to be a passive decay heat removal. Here, 7
you've got a combination of passive/active/quasi-8 active systems, maybe depending on what design 9
approaches we see presented, but you would really hear 10 about containment bypass, which is not in the MHTGR 11 designs. That, essentially, is confinement and 12 reliance, mainly, on the functional containment. So, 13 I think that's interesting from my vantage point, 14 looking at how they picked and choose, and how you 15 reviewed their use of the Reg Guide.
16 That would be the two areas I would zero-17 in on for this particular design. And it begs the 18 question, like Charlie was asking, you know, which 19 systems are active in terms of which ones might need 20 electric power, or will they fail safe, so to speak, 21 without power?
22 MEMBER PETTI: Well, it just seems to me 23 that there's sort of a body of knowledge of MHTGR and 24 there's a body of knowledge to fast sodium systems.
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108 And this kind of sort of puts them together. There's 1
an intersection, if you will. So, when one thinks 2
about accident response, you know, there may be 3
something there that, when you look at them in 4
accident space, something new comes up that you 5
wouldn't necessarily see looking at them each 6
separately. So, it's just something that, when you 7
get into the details, you have to be looking for.
8 CHAIR BIER: Additional questions or 9
comments from members or consultants? Anybody on the 10 line?
11 If not, then we are going to take comments 12 from the public somewhat earlier than is indicated on 13 the agenda.
14 We'll wait another 30 seconds or so, in 15 case anybody is trying to unmute.
16 (No response.)
17 CHAIR BIER: Okay. It sounds like we have 18 no public comments for today.
19 So, at this point, we have time for member 20 discussion. And I forget if that should be public or 21 not public.
22 MEMBER REMPE: We can go off the record 23 until regular order. But hope someone will show up 24 tomorrow at 8:30 for us.
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109 CHAIR BIER: Okay, you got that message, 1
Court Reporter? I don't believe there is a need for 2
a closed session. So we will see you at 8:30 in the 3
morning tomorrow, or whoever it is. Thank you.
4 (Whereupon, the above-entitled matter went 5
off the record at 3:39 p.m.)
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
1 General Atomics Electromagnetic Systems Fast Modular Reactor Principal Design Criteria Presented To: Advisory Committee on Reactor Safeguards May 2, 2023 Prepared By:
John Bolin (GA-EMS)
2 Principal Design Criteria Adapted From NRC Guidance
- Regulatory Guide (RG) 1.232 established guidance for developing PDC for non-light-water reactors by modifying / supplementing 10 CFR 50, Appendix A, General Design Criteria (GDC) in three categories:
- Sodium-cooled fast reactors (SFR-DC)
- Modular high-temperature gas-cooled reactors (MHTGR-DC)
- Design-neutral advanced reactors (ARDC)
- ARDC and MHTGR-DC used as starting point
- NRC rationale for adaptation of GDC modified for application to FMR-DC
3 I. Overall Requirements - FMR DC 1 - 5
- FMR-DC 1: Quality standards and records: Same as GDC
- FMR-DC 2: Design bases for protection against natural phenomena:
Same as GDC
- FMR-DC 3: Fire protection: Same as ARDC
- FMR-DC 4: Environmental and dynamic effects design bases:
Modified from MHTGR-DC
- missiles changed to missiles originating both inside and outside the reactor helium pressure boundary to cover turbomachinery
- FMR-DC 5: Sharing of structures, systems, and components:
Same as GDC 4 II. Multiple Barriers - FMR-DC 10 - 19
- FMR-DC 10: Reactor design: Fuel design using SiGA cladding functions like LWRs so SAFDL terminology used here and in other FMR-DCs
- FMR-DC 11, 13 - 15, 17 - 18: Same as ARDC and MHTGR-DC with minor terminology changes
- FMR-DC 12: Suppression of reactor power oscillations: The word structures added to address reflectors. The word coolant was deleted.
- FMR-DC 16: Containment design: Same as SFR-DC because FMR uses low-leakage, pressure-retaining containment
5 III. Reactivity Control - FMR-DC 20 - 29
- FMR-DC 20 - 24: Same as GDC
- FMR-DC 25: Protection system requirements for reactivity control malfunctions: Same as ARDC with minor terminology changes
- FMR-DC 28: Reactivity limits: Same as MHTGR-DC
- FMR-DC 29: Protection against anticipated operational occurrences:
Same as GDC 6 IV. Fluid Systems - FMR-DC 30 - 46
- FMR-DC 30 - 33: Same as MHTGR-DC
- FMR-DC 34: Residual heat removal: Similar to MHTGR-DC. Both active non-safety-related and passive safety-related systems available to remove residual heat. Incorporates requirements in GDC 35.
- FMR-DC 36 and 37: Same as MHTGR-DC
- FMR-DC 38 - 41: Same as ARDC
- FMR-DC 42: Inspection of containment atmosphere cleanup systems:
Same as GDC
- FMR-DC 43, 45, 46: Same as ARDC
- FMR-DC 44: Same as MHTGR-DC
7 V. Reactor Containment and VI. Fuel and Radioactivity Control V. Reactor Containment - FMR-DC 50-57
- FMR-DC 50 - 53: Same as ARDC
- FMR-DC 54: Same as SFR-DC
- FMR-DC 55 - 57: Same as ARDC with minor terminology changes VI. Fuel and Radioactivity Control - FMR-DC 60 - 64
- FMR-DC 60, 62, 63: Same as GDC
- FMR-DC 61: Same as ARDC
8 Acknowledgements This work was supported by the U.S. Department of Energy - Office of Nuclear Energy under Contract Number DE-NE0009052 for Advanced Reactor Concepts-20 (ARC-20).
General Atomics - Electromagnetic Systems Fast Modular Reactor Principal Design Criteria Samuel Cuadrado de Jesus, NRR/DANU Reed Anzalone, NRR/DANU Sheila Ray, NRR/DEX Steve Jones, NRR/DANU
Agenda
- General Atomics - Electromagnetic Systems (GA-EMS) Fast Modular Reactor (FMR) pre-application engagement
- GA-EMS FMR design features
- PDC guidance
- General Design Criteria (GDC)
- GA-EMS PDC development approach
- Fast modular reactor design criteria (FMR-DC)
- Impacts of key design choices on PDCs
- FMR-DC overview
- Safety evaluation (SE) conclusions
GA-EMS FMR Pre-Application Engagement Documents Submitted Submittal Document Review Status 03/2022 Pre-Application Regulatory Engagement Plan N/A 06/2022 PDC TR Draft SE issued 02/2023 Fuel Qualification Plan Technical Report Under review (white paper) 04/2023 Quality Assurance Program TR Pending acceptance determination Documents Expected Submittal Document 06/2023 Mechanistic Source Term Technical Report 12/2023 LBE Selection White Paper 05/2024 Safety Approach and Mini-PRA White Paper 05/2024 Safety Classification White Paper
- FMR demonstration expected by 2030 and deployment by mid-2030s
GA-EMS FMR PDC TR Review Timeline
- Submitted 06/06/22 (ML22154A555)
- Accepted 07/07/22 (ML22181B173)
- Requests for Additional Information (RAIs) issued 10/5/22 (ML22321A310)
- RAI response received 11/7/22 (ML22311A472)
- Revision 2 of TR transmitted 01/05/23 (ML23005A292)
- Draft SE issued 03/17/23 (ML23076A196)
GA-EMS FMR Design Features Source: REP, ML22087A510
GA-EMS FMR Design Features Source: TR, ML22154A556
GA-EMS FMR Key Design Features Feature Design Fuel UO2 pellets in silicon carbide fuel pins Core arrangement Pins in triangular pitch arranged into hexagonal bundles Neutron spectrum Fast Coolant Helium Power conversion system Gas turbine on primary coolant Residual heat removal Reactor vessel cooling system (water-fed, gravity-driven passive system)
Containment Leak-tight containment building
PDC Guidance - 10 CFR 50 Appendix A GDC The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
These General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The General Design Criteria are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units.
PDC Guidance - RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors
- Issued April 2018 (ACRS letter March 2018)
- Documents three sets of acceptable PDCs:
- Advanced reactor DC (ARDC) - generic, technology inclusive*
- Sodium-cooled fast reactor DC (SFR-DC) - sodium-cooled fast reactors (e.g.,
PRISM)
- Modular high temperature gas-cooled reactor DC (MHTGR-DC) -
TRISO-fueled, helium-cooled, graphite-moderated HTGR
- For sodium/lead/gas-cooled fast reactors, modular high temperature gas reactors, fluoride high-temperature reactors, and molten salt reactors
GA-EMS Approach to PDC Development
- Start with ARDC, considering underlying safety basis
- If ARDC not fully applicable, assess SFR-DC and MHTGR-DC for direct adoption
- If SFR-DC or MHTGR-DC not directly applicable, apply DC that is most representative of FMR
- Adapt or refine selected DC
Key Design Feature Effects on PDCs Feature Design Effect on PDCs Fuel UO2 pellets in silicon carbide fuel pins Use of specified acceptable fuel design limits (SAFDLs) instead of specified acceptable system radionuclide release design limits (SARRDLs)
Core arrangement Pins in triangular pitch arranged into hexagonal bundles Neutron spectrum Fast Consider effect of structures on reactivity feedback Coolant Helium Removal of coolant inventory control considerations consistent with MHTGR; use of reactor helium pressure boundary in lieu of reactor coolant pressure boundary Power conversion system Gas turbine on primary coolant No particular effect Residual heat removal Reactor vessel cooling system (water-fed, gravity-driven passive system)
Adoption of MHTGR passive residual heat removal PDCs Containment Leak-tight containment building Adoption of containment PDCs
FMR-DC - I. Overall Requirements Criterion Title Basis PDC Modified?
1 Quality standards and records.
ARDC N
2 Design bases for protection against natural phenomena.
ARDC N
3 Fire protection.
ARDC N
4 Environmental and dynamic effects design bases.
MHTGR-DC N
5 Sharing of structures, systems, and components ARDC N
FMR-DC - II. Multiple Barriers Criterion Title Basis PDC Modified?
10 Reactor design.
ARDC Y - uses "heat removal" instead of "coolant" 11 Reactor inherent protection.
ARDC N
12 Suppression of reactor power oscillations.
ARDC Y - removes "coolant" 13 Instrumentation and control.
ARDC Y - uses "helium pressure boundary" instead of "reactor coolant boundary" 14 Reactor helium pressure boundary.
MHTGR-DC N
15 Reactor helium pressure boundary design.
MHTGR-DC N
16 Containment design.
SFR-DC N
17 Electric power systems.
MHTGR-DC Y - uses SAFDLs instead of SARRDLs 18 Inspection and testing of electric power systems.
ARDC N
19 Control room.
MHTGR-DC N
FMR-DC - III. Reactivity Control Criterion Title Basis PDC Modified?
20 Protection system functions ARDC N
21 Protection system testability and reliability.
ARDC N
22 Protection system independence.
ARDC N
23 Protection system failure modes.
ARDC N
24 Separation of protection and control systems.
ARDC N
25 Protection system requirements for reactivity control malfunctions.
ARDC N
26 Reactivity control systems.
ARDC Y - includes effects of xenon 27
[None - incorporated into 26 consistent with RG 1.232]
N/A N/A 28 Reactivity limits.
MHTGR-DC N
29 Protection against anticipated operational occurrences.
ARDC N
FMR-DC - IV. Fluid Systems (1)
Criterion Title Basis PDC Modified?
30 Quality of reactor helium pressure boundary.
MHTGR-DC N
31 Fracture prevention of reactor helium pressure boundary. MHTGR-DC N
32 Inspection of reactor helium pressure boundary MHTGR-DC N
33
[None - not applicable consistent with MHTGR-DC]
N/A N/A 34 Residual heat removal.
MHTGR-DC Y - includes both passive and active systems 35
[None - not applicable consistent with MHTGR-DC]
N/A N/A 36 Inspection of passive residual heat removal system.
MHTGR-DC N
37 Testing of residual heat removal system.
MHTGR-DC Y - includes both passive and active systems 38 Containment heat removal.
ARDC N
39 Inspection of containment heat removal system.
ARDC N
FMR-DC - IV. Fluid Systems (2)
Criterion Title Basis PDC Modified?
40 Testing of containment heat removal system.
ARDC N
41 Containment atmosphere cleanup.
ARDC N
42 Inspection of containment atmosphere cleanup systems. ARDC N
43 Testing of containment atmosphere cleanup systems.
ARDC N
44 Structural and equipment cooling.
ARDC N
45 Inspection of structural and equipment cooling systems.
ARDC N
46 Testing of structural and equipment cooling systems.
ARDC N
FMR-DC - V. Reactor Containment Criterion Title Basis PDC Modified?
50 Containment design basis.
ARDC N
51 Fracture prevention of containment pressure boundary.
ARDC N
52 Capability for containment leakage rate testing.
ARDC N
53 Provisions for containment testing and inspection.
ARDC N
54 Piping systems penetrating containment.
SFR-DC Y - removes "reactor" 55 Reactor helium pressure boundary penetrating containment.
ARDC Y - uses "helium pressure boundary" instead of "reactor coolant boundary" 56 Containment isolation.
ARDC N
57 Closed system isolation valves.
ARDC Y - uses "helium pressure boundary" instead of "reactor coolant boundary"
FMR-DC - VI. Fuel and Reactivity Control Criterion Title Basis PDC Modified?
60 Control of releases of radioactive materials to the environment.
ARDC N
61 Fuel storage and handling and radioactivity control.
ARDC N
62 Prevention of criticality in fuel storage and handling.
ARDC N
63 Monitoring fuel and waste storage.
ARDC N
64 Monitoring radioactivity releases.
ARDC N
Safety Evaluation Conclusions
- GA-EMS appropriately considered RG 1.232 and developed a sufficient set of PDCs appropriate for establishing requirements for the FMR design.
- PDCs establish the necessary design, fabrication, construction, testing, and performance design criteria for safety-significant SSCs to provide reasonable assurance that an FMR could be operated without undue risk to the health and safety of the public. (10 CFR 50 App A)
- This TR can be used by future FMR applicants, but if the reactor design differs from that discussed in the TR use of the PDCs in the TR must be justified.
FMR-DC Summary
- Directly adopted from RG 1.232
- From ARDC: FMR-DC 1, 2, 3, 5, 11, 18, 20, 21, 22, 23, 24, 25, 29, 38, 39, 40, 41, 42, 43, 44, 45, 46, 50, 51, 52, 53, 60, 61, 62, 63, 64
- From SFR-DC: FMR-DC 16
- From MHTGR-DC: FMR-DC 4, 14, 15, 19, 28, 30, 31, 32, 36
- Modified from RG 1.232
- FMR-DC 10 (ARDC 10), 12 (ARDC 12), 13 (ARDC 13), 17 (MHTGR-DC 17), 26 (ARDC 26), 34 (MHTGR-DC 34), 37 (MHTGR-DC 37), 54 (SFR-DC 54), 55 (ARDC 55), 57 (ARDC 57)
FMR-DC Modified from RG 1.232 ARDC 10 FMR-DC 10 Reactor design.
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Reactor design.
The reactor core and associated coolant heat removal, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Basis: Helium inventory control is not necessary to meet SAFDLs due to reactor system design; consistent with MHTGR-DC (which use SARRDLs instead) and other FMR-DC
FMR-DC Modified from RG 1.232 ARDC 12 FMR-DC 12 Suppression of reactor power oscillations.
The reactor core; associated structures; and associated coolant, control, and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Suppression of reactor power oscillations.
The reactor core;, associated structures;,
and associated coolant, control, and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Basis: Helium coolant does not have a significant effect on reactivity for the FMR
FMR-DC Modified from RG 1.232 ARDC 13 FMR-DC 13 Instrumentation and control.
Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate to ensure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Instrumentation and control.
Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate, to ensure adequate safety, including those variables and systems that can affect the fission process, and the integrity of the reactor core, the reactor coolant helium pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Basis: More appropriate to say reactor helium pressure boundary than reactor coolant boundary for FMR, consistent with MHTGR-DC and other FMR-DC
FMR-DC Modified from RG 1.232 ARDC 26 FMR-DC 26 Reactivity control systems.
A minimum of two reactivity control systems or means shall provide:
(1) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the design limits for the fission product barriers are not exceeded and safe shutdown is achieved and maintained during normal operation, including anticipated operational occurrences.
(2) A means which is independent and diverse from the other(s), shall be capable of controlling the rate of reactivity changes resulting from planned, normal power changes to assure that the design limits for the fission product barriers are not exceeded.
(3) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the capability to cool the core is maintained and a means of shutting down the reactor and maintaining, at a minimum, a safe shutdown condition following a postulated accident.
(4) A means for holding the reactor shutdown under conditions which allow for interventions such as fuel loading, inspection and repair shall be provided.
Reactivity control systems.
A minimum of two reactivity control systems or means shall provide:
(1) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the design limits for the fission product barriers are not exceeded and safe shutdown is achieved and maintained during normal operation, including anticipated operational occurrences.
(2) A means which is independent and diverse from the other(s), shall be capable of controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure that the design limits for the fission product barriers are not exceeded.
(3) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the capability to cool the core is maintained and a means of shutting down the reactor and maintaining, at a minimum, a safe shutdown condition following a postulated accident.
(4) A means for holding the reactor shutdown under conditions which allow for interventions such as fuel loading, inspection and repair shall be provided.
Basis: GDC 26 includes explicit consideration of Xe burnout; while Xe is not expected to be a significant reactivity contributor in the FMR it is not incorrect to explicitly include it
FMR-DC Modified from RG 1.232 MHTGR-DC 34 FMR-DC 34 Passive residual heat removal.
A passive system to remove residual heat shall be provided. For normal operations and anticipated operational occurrences, the system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core to an ultimate heat sink at a rate such that specified acceptable system radionuclide release design limits and the design conditions of the reactor helium pressure boundary are not exceeded.
During postulated accidents, the system safety function shall provide effective cooling.
Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure the system safety function can be accomplished, assuming a single failure.
Passive rResidual heat removal.
A passive sSystem(s) to remove residual heat shall be provided. For normal operations and anticipated operational occurrences, the system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core to an ultimate heat sink at a rate such that specified acceptable system radionuclide release fuel design limits and the design conditions of the reactor helium pressure boundary are not exceeded.
During postulated accidents, the system safety function shall provide effective core cooling.
Suitable redundancy in components and features and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure the system safety function can be accomplished, assuming a single failure.
Basis: The MHTGR included a passive residual heat removal (RHR) system because of the low core power density. FMR has multiple RHR systems including active non-safety-related systems and passive safety-related systems, and the DC should be broad enough to apply to all of them.
FMR-DC Modified from RG 1.232 MHTGR-DC 37 FMR-DC 37 Testing of passive residual heat removal system.
The passive residual heat removal system shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including associated systems, for AOO or postulated accident decay heat removal to the ultimate heat sink and, if applicable, any system(s) necessary to transition from active normal operation to passive mode.
Testing of passive residual heat removal system.
The passive residual heat removal system(s) shall be designed to permit appropriate periodic functional testing to ensure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the system components, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including associated systems, for AOO or postulated accident decay heat removal to the ultimate heat sink and, if applicable, any system(s) necessary to transition from active normal operation to passive mode.
Basis: The MHTGR included a passive residual heat removal (RHR) system because of the low core power density. FMR has multiple RHR systems including active non-safety-related systems and passive safety-related systems, and the DC should be broad enough to apply to all of them (same as FMR-DC 34).
FMR-DC Modified from RG 1.232 SFR-DC 54 FMR-DC 54 Piping systems penetrating containment.
Piping systems penetrating the reactor containment structure shall be provided with leak detection, isolation, and containment capabilities that have redundancy, reliability, and performance capabilities necessary to perform the containment safety function and that reflect the importance to safety of preventing radioactivity releases from containment through these piping systems. Such piping systems shall be designed with the capability to verify, by testing, the operational readiness of any isolation valves and associated apparatus periodically and to confirm that valve leakage is within acceptable limits.
Piping systems penetrating containment.
Piping systems penetrating the reactor containment structure shall be provided with leak detection, isolation, and containment capabilities that have redundancy, reliability, and performance capabilities necessary to perform the containment safety function and that reflect the importance to safety of preventing radioactivity releases from containment through these piping systems. Such piping systems shall be designed with the capability to verify, by testing, the operational readiness of any isolation valves and associated apparatus periodically and to confirm that valve leakage is within acceptable limits.
Basis: There are other major SSCs other than just the reactor within containment (e.g.,
the power conversion system) so it is appropriate to remove the word reactor
FMR-DC Modified from RG 1.232 ARDC 55 FMR-DC 55 Reactor coolant boundary penetrating containment.
Each line that is part of the reactor coolant boundary and that penetrates the containment structure shall be provided with containment isolation valves, as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
Reactor coolant helium pressure boundary penetrating containment.
Each line that is part of the reactor coolant helium pressure boundary and that penetrates the reactor containment structure shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:
Basis: More appropriate to say reactor helium pressure boundary than reactor coolant boundary for FMR, consistent with MHTGR-DC and other FMR-DC
FMR-DC Modified from RG 1.232 ARDC 57 FMR-DC 57 Closed system isolation valves.
Each line that penetrates the containment structure and is neither part of the reactor coolant boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve, unless it can be demonstrated that the containment safety function can be met without an isolation valve and assuming failure of a single active component. The isolation valve, if required, shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.
Closed system isolation valves.
Each line that penetrates the containment structure and is neither part of the reactor coolant helium pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve unless it can be demonstrated that the containment safety function can be met without an isolation valve and assuming failure of a single active component. The isolation valve, if required, shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.
Basis: More appropriate to say reactor helium pressure boundary than reactor coolant boundary for FMR, consistent with MHTGR-DC and other FMR-DC