ML23123A029
| ML23123A029 | |
| Person / Time | |
|---|---|
| Issue date: | 04/06/2023 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-2344 | |
| Download: ML23123A029 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Open Session Location: teleconference Date: 04-06-2023 Work Order No.: NRC-2344 Pages 1-94 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 704TH MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 OPEN SESSION 9 + + + + +
10 THURSDAY 11 APRIL 6, 2023 12 + + + + +
13 The Advisory Committee met via hybrid In-14 Person and Video-Teleconference, at 8:30 a.m. EDT, Joy 15 L. Rempe, Chairman, presiding.
16 17 COMMITTEE MEMBERS:
18 JOY L. REMPE, Chairman 19 WALTER L. KIRCHNER, Vice Chairman 20 DAVID A. PETTI, Member-at-Large 21 RONALD G. BALLINGER, Member 22 CHARLES H. BROWN, JR., Member 23 VICKI M. BIER, Member 24 VESNA B. DIMITRIJEVIC, Member 25 GREGORY H. HALNON, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 1 JOSE MARCH-LEUBA, Member 2 MATTHEW W. SUNSERI, Member 3
4 ACRS CONSULTANT:
5 DENNIS BLEY 6 STEPHEN SCHULTZ 7
8 DESIGNATED FEDERAL OFFICIAL:
9 CHRISTINA ANTONESCU 10 LARRY BURKHART 11 12 ALSO PRESENT:
13 JOE ASHCRAFT, NRR 14 ERIC BENNER, NRR 15 GILBERTO BLAS RODRIGUEZ, NRR 16 SAMIR DARBALI, NRR 17 WILLIAM JESSUP, NRR 18 CHRIS LEVESQUE, TerraPower 19 ED LYMAN, Public Participant 20 KHOI NGUYEN, NRR 21 TARA NEIDER, TerraPower 22 JASON PAIGE, NRR 23 RYAN SPRENGEL, TerraPower 24 RICHARD STATTEL, NRR 25 DINESH TANEJA, NRR NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 1 MARK WERNER, TerraPower 2 ERIC WILLIAMS, TerraPower 3 GEORGE WILSON, TerraPower 4 BRIAN YIP, NSIR 5
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
4 1 P R O C E E D I N G S 2 8:30 a.m.
3 CHAIR REMPE: Good morning. It is 8:30 on 4 the East Coast, and this meeting will now come to 5 order. This is the second day of the 704th Meeting of 6 the Advisory Committee on Reactor Safeguards.
7 I'm Joy Rempe, Chairman of the ACRS.
8 Other members in attendance are Ron Ballinger, Vicki 9 Bier, Charles Brown, Vesna Dimitrijevic, Greg Halnon, 10 Walt Kirchner, Jose March-Leuba, Dave Petti, and Matt 11 Sunseri. We do have a quorum.
12 Similar to yesterday, the Committee is 13 meeting in person and virtually. A communications 14 channel has been opened to allow members of the public 15 to monitor the Committee discussion. Mr. Larry 16 Burkhart is the Designated Federal Officer for today's 17 meeting.
18 During today's meeting, the Committee will 19 consider the following topics: planning and procedures 20 session and Commission meeting preparation, TerraPower 21 and Natrium reactor design overview and 3D model 22 walkthrough. I note that portions of our discussions 23 on both of these items may be closed.
24 The transcript of the open portions of the 25 discussion on topic two is being kept, and it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 1 requested that speakers identify themselves and speak 2 with sufficient clarity and volume so they can be 3 readily heard. Additionally, participants should mute 4 themselves when they're not speaking.
5 At this time, I'd like to ask other 6 members if they have any opening remarks. Not hearing 7 any, then I'd like to go off the record at this time 8 before we start with topic one today. And I'd also 9 ask the court reporter to return at 1 p.m. for our 10 discussion on topic two.
11 (Whereupon, the above-entitled matter went 12 off the record at 8:31 a.m. and then went back on the 13 record at 1:00 p.m.)
14 CHAIR REMPE: Okay. It's 1 p.m. on the 15 East Coast, and we are going to reconvene, and I'm 16 going to ask Member Kirchner to lead us through the 17 second topic for today. Walt.
18 VICE CHAIR KIRCHNER: Okay. We have 19 guests here today with us from TerraPower, and they 20 are going to give us an informational presentation on 21 the Natrium design they are developing and give us an 22 overview. I'd ask Chris Levesque to perhaps somewhere 23 through this address the status of their engagement 24 with the staff, which is useful information for us 25 when we look at our longer-range planning going NEAL 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 1 forward.
2 So with that, Chris, I'm just going to 3 turn it over to you to introduce your team. And thank 4 you for being here. We like the engagement in person, 5 so, hopefully, you'll like the engagement with us in 6 person, as well. Go ahead.
7 MR. LEVESQUE: Thank you, Madam Chair, Mr.
8 Vice Chair, and thanks to the entire Advisory 9 Committee on Reactor Safeguards at this time. We look 10 forward to the engagement today and to multiple 11 engagements with you in the future as we go forward 12 with the Natrium licensing process.
13 I'm joined by Tara Neider, who is the 14 project director for the Natrium project, Senior Vice 15 President and project director at TerraPower. She 16 oversees the entire Natrium project, including the 800 17 design engineers who are working on the project today 18 in the procurement and instruction processes, as well.
19 I think I might defer further intros from 20 the TerraPower team because we'll have multiple people 21 coming up today, and we'll save those intros for when 22 those subject matter experts come in.
23 I wanted to begin by sharing, you know, 24 the common value, the common understanding that we at 25 TerraPower feel towards reactor safety. We believe NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 1 the Natrium design, you know, addresses the NRC's 2 requirements and the engineering requirements that are 3 needed for reactor safety. It is a Gen IV reactor, 4 which we expect Gen IV reactors to have even higher 5 margins for safety and we'll utilize excessively safe 6 and inherent systems to achieve safety.
7 MR. NGUYEN: Excuse me. Could you 8 identify yourself for the court reporter? Thank you.
9 MR. LEVESQUE: Yes. My name is Chris 10 Levesque, President and CEO of TerraPower.
11 We're also looking forward, in addition to 12 the engagement with the ACRS, we're closely monitoring 13 the development of the Advanced Reactor Content of 14 Application Project. We know, following the Nuclear 15 Energy Innovation and Modernization Act, there's 16 various developments going on in regulation, so we're 17 watching these closely. But we are making this first 18 application under 10 CFR 50, an established statute.
19 We're also following the NRC's guidance on 20 the pre-application engagement, and we have 21 considerable pre-application engagement already behind 22 us with 39 meetings in pre-application, 5 topical 23 reports, and 3 white papers. And that engagement with 24 the NRC has been very constructive. We look forward 25 to engagements like this with the ACRS and engagements NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 1 with the NRC in the pre-application review as a way of 2 improving our eventual construction permit application 3 and operating license application.
4 Knowing that Natrium is a Gen IV design 5 and most of the experience in the U.S. is in the light 6 water arena, we're doing as much as we can to 7 familiarize the staff and other stakeholders on sodium 8 fast reactors. Best example of that is the multiple 9 training sessions we've given for NRC staff on sodium 10 fast reactors and the attributes of Natrium.
11 I want to finish by noting that this first 12 Natrium plant, which will be built in the state of 13 Wyoming at a retiring coal plant, is being built in 14 conjunction with a very important public-private 15 partnership, the Advanced Reactor Demonstration 16 Program. Both TerraPower and X-energy were recipients 17 of the ARDP Demonstration Project Award. That award 18 involves the Department of Energy funding half of the 19 costs of these first projects, and that's intended to 20 help us overcome many first-time costs associated with 21 the first license, the first design, learning curve, 22 frankly, because the U.S. has somewhat fallen out of 23 experience with new builds.
24 So the ARDP, we feel, is a very, very 25 important national program. We're proud to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 1 recipients of that program, and a very important part 2 of that program is that the first owner of the Natrium 3 nuclear power plant will be a commercial owner.
4 So we believe that public-private 5 partnerships are very important. We're competing on 6 the world stage with Chinese and Russian-state owned 7 entities. We believe that American innovation and 8 American public-private partnerships are really the 9 way to compete with the state-owned entities, and we 10 look forward to engagement with all of our government 11 stakeholders.
12 I'm going to now hand it off to Tara 13 Neider, again, our Senior Vice President and project 14 director for Natrium.
15 CHAIR REMPE: So I don't know if this is 16 your first time at ACRS, but we like to ask questions.
17 MR. LEVESQUE: Please.
18 CHAIR REMPE: Are we allowed to ask you 19 questions? I know you've got a meeting you've got to 20 go --
21 MR. LEVESQUE: Absolutely. No, we look 22 forward to that. Yes.
23 CHAIR REMPE: I'm just curious. I mean, 24 you've mentioned that you've got five topical reports 25 in the queue and you mentioned that you look forward NEAL 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 1 to future engagements with ACRS. Do you have a time 2 frame when you think you'll be back before ACRS?
3 MR. LEVESQUE: Well, we'll share, in 4 further reports, we'll share our plans for our 5 construction permit application, which will be early 6 next year. We would plan multiple meetings. We 7 certainly don't think it's too late or too early to be 8 engaging you because we have a fully mobilized 9 project. I know there's many projects that you can 10 read about everyday in the U.S., but I would hazard to 11 say maybe none of them have 800 design engineers 12 working on a design, soil borings at the site already 13 complete. I mean, this project is quite advanced and, 14 given the construction permit application early next 15 year, we think this is a great time for the first ACRS 16 meeting.
17 George, Ryan, any --
18 MR. WILSON: This is George Wilson. I'm 19 the VP of Regulatory Affairs for TerraPower. Right 20 now, Billy Jessup is in the room, we have scheduled 21 meetings --
22 MR. BLEY: Can you talk more into the 23 microphone?
24 CHAIR REMPE: You need to be closer to the 25 mike for the people online. I'm sorry.
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11 1 MR. WILSON: This is George Wilson, 2 TerraPower, VP of Regulatory Affairs. I know that we 3 have some topical report meetings lined up with the 4 ACRS starting in August, this August. But we wanted 5 to come in, give an overview now, so that you guys 6 could potentially ask questions and then we can, you 7 know, as we go from there.
8 CHAIR REMPE: I'd caution you that it's 9 not showing up, we have our schedule that we went 10 through earlier today, and it's not on the schedule.
11 So please work with our staff to make sure that it is 12 as time that members know in advance, and they can 13 make their plans accordingly if you're going to be 14 coming in in August. That's why I asked --
15 MR. WILSON: That was the feedback that we 16 had gotten from NRR. I think Branch Chief Billy 17 Jessup is here, if he wants to add some more.
18 CHAIR REMPE: You can just stand and talk.
19 We've gone with new technology. And be sure and say 20 your name.
21 MR. JESSUP: Good afternoon. This is Bill 22 Jessup, Chief of Advanced Reactor Licensing Branch 1, 23 NRR. Our project managers are working very closely 24 with ACRS staff as we progress through the review of 25 the various topical reports and white papers, as well.
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12 1 I know the ACRS is not looking at white papers. We 2 are working very closely with ACRS staff to schedule 3 forthcoming meetings that have been --
4 CHAIR REMPE: It's not on the August 5 schedule. If we're going to meet in August, be sure 6 and get it on the schedule, but that helps to know.
7 And at some point, I know some of this proprietary, 8 but some of your, if you can tell us which report it 9 is that's going to be coming in in August, that would 10 be helpful for us, too, to know. I don't know if you 11 want to say it on the record here.
12 MR. WILSON: I think the NRC staff, we 13 have a -- this is George Wilson again. We have a 14 topical report. Our reactor design is a little bit 15 different. We have a topical report on the nuclear 16 island, energy island interfaces. Since we can 17 actually operate the turbine, it's not dependent, we 18 can operate the turbine at, like, 100-percent power 19 and the reactor at 10-percent power. They're not 20 linked like a light water reactor, so, based on that, 21 there are certain regulations that will not apply to 22 the Natrium design. So that is a topical report right 23 now the NRC staff has completed the audit on and 24 they're finishing up their review, and I know that 25 will be one that's coming to ACRS.
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13 1 We submitted principal design criteria.
2 That will be another one that the NRC has just started 3 reviewing. So those are ones that will be coming up 4 to the ACRS first.
5 CHAIR REMPE: Okay. I did look over your 6 regulatory engagement plan, but I just couldn't get a 7 clue of what's coming in and when. So thank you.
8 MS. NEIDER: As was mentioned, I'm Tara 9 Neider. I'm the project --
10 MEMBER PETTI: Microphone, please.
11 MS. NEIDER: Oh, okay. As Chris 12 mentioned, I'm Tara Neider. I'm the project director 13 for Natrium, and I wanted to talk a little bit about 14 the Kemmerer site that we're selecting to build this 15 plant on and then discuss a little bit about the team 16 that we've put together, which I think is a pretty 17 world class team.
18 As Chris mentioned, we are building the 19 Natrium reactor at the site of a coal plant that is 20 scheduled to shut down. And I can tell you that when 21 we did our tour of various places to put the plant in 22 Wyoming, this town, actually there's two towns, 23 Diamondville and Kemmerer, they really welcomed us 24 with open arms. It's really important for this 25 community to have something to replace the work that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 1 was done at both the coal mine, you know, which feeds 2 the coal plant, and then the coal plant. So it's been 3 a very, very cooperative agreement we've had with the 4 town. However, it is a small town, so we do have to 5 make sure that we can be prepared for the number of 6 people that we have which will be coming during the 7 construction to build the plant.
8 The plant is a sodium fast reactor, and 9 it's a pool type reactor, and most of the safety-10 related systems are underground. Mark and Eric will 11 be talking later about the details of that design.
12 But what really puts us apart from others is that the 13 power generated from Natrium is dispatchable. Between 14 the nuclear island and the energy island, nuclear 15 island can run all the time, the energy island can 16 actually go up and down in power as the power is 17 needed on the grid.
18 Our team is, as I said, a world class 19 team. Our design partner is GE-Hitachi. They have 20 very strong experience in sodium fast reactors and, 21 obviously, commercial BWR plants. And we also have 22 Bechtel. Bechtel is doing the construction, and 23 they're also responsible for the design of the energy 24 island and some of the civil works on the nuclear 25 island.
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15 1 We also have a lot of other entities that 2 are part of our team. Specifically, we've engaged 3 with a number of nuclear utilities to provide us 4 guidance on the operations piece and make sure that 5 both operations and maintenance is considered in our 6 design. Those include Duke Energy and Energy 7 Northwest, as well as Pacific Corp, who will be the 8 owner/operator of the plant eventually. Pacific Corp 9 or, actually, a subsidiary of Rocky Mountain Power is 10 the utility in that area, and they will eventually own 11 and operate the plant, so they are part of our team 12 moving forward.
13 We also have a number of national labs 14 that are doing various support with regard to modes 15 and methods and also testing of our advanced fuel for 16 this reactor. And we have a number of other entities, 17 as well, a number of universities. The University of 18 Wisconsin, Oregon, and also NC State is part of our 19 team. So we really feel that we've put together a 20 pretty top-notch team to be able to move forward.
21 In addition to our teammates, we are 22 trying to make sure that our plant meets the needs of 23 the commercial market. And as a result of that, we've 24 formed what we call the Natrium utility advisory 25 group, but we also now call it users because there's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 1 a number of companies in the advisory committee that 2 are not utilities at all. They're industry 3 participants. And so those people we meet with on a 4 regular basis, and they provide us guidance and take 5 a look at how we're going through things. And I think 6 we're up to about 30 people, 30 companies on that 7 advisory group, so it's very positive.
8 We just joined INPO as just a supplier at 9 this point, but we think that having joined INPO will 10 actually provide us a lot of experience, especially 11 the lessons learned from the industry going forward.
12 And then, finally, we also have a memorandum of 13 understanding with JAEA, the Japanese Atomic Energy 14 Agency, and Mitsubishi. The reason that we have that 15 memorandum in place is that they will be providing 16 their own sodium fast reactors, but they also have a 17 fair amount of experience running Joyo and Monju in 18 Japan, so we really felt that their guidance would 19 help us with this plant. And where they've helped us 20 if they've participated in some of our design reviews 21 going forward.
22 So those are my main points I wanted to 23 make today, and I think, at this point, we'll ask if 24 you have any questions or comments before we get into 25 the technical side.
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17 1 MR. BLEY: This is Dennis Bley. You've 2 said you joined INPO. Are you considering being 3 involved with FLEX and SAFER, or does that just not 4 match up with you at all?
5 MS. NEIDER: We are considering FLEX and 6 SAFER, yes. Not SAFER. FLEX, I know, we are 7 considering a FLEX building, but it's going to depend 8 on where our analysis comes up with things.
9 MR. BLEY: Thank you.
10 VICE CHAIR KIRCHNER: Maybe we'll get into 11 this as we go forward, but, just at a very high level, 12 just the first order, is this an evolution of a PRISM 13 design that GE had been advocating back in the mid -
14 late 80s time frame?
15 MS. NEIDER: There are certain elements of 16 the PRISM design. We kind of combined the TerraPower 17 traveling wave reactor and the GE PRISM reactor to 18 develop the Natrium reactor. The reason, you know, 19 the main change from those two is the size, but we did 20 find some things on the various components that we did 21 have to do differently just because --
22 VICE CHAIR KIRCHNER: When we get into 23 closed session, maybe just keep this in mind. I'd 24 like to ask some questions about what you learned from 25 the PRISM design. In particular, the NRC back in that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
18 1 time frame, actually, a CER on the PRISM reactor, so 2 it did get a fair amount of attention --
3 MS. NEIDER: Yes. And that safety 4 evaluation report has been combed over by our 5 engineers, and there were certain open items, like 6 transient testing, that was required, and we've added 7 that to our program. So we do believe those things 8 are covered.
9 VICE CHAIR KIRCHNER: Good. Is this going 10 to be the end of your open presentation, or are you 11 going to shift gears here?
12 MR. WILLIAMS: We haven't actually started 13 our open presentation.
14 (Laughter.)
15 VICE CHAIR KIRCHNER: Well, it's not 16 unusual at ACRS to not get beyond the first --
17 (Laughter.)
18 VICE CHAIR KIRCHNER: -- so I'm just 19 checking. Okay. Go ahead.
20 MR. WILLIAMS: Okay. I'll introduce 21 myself real quick. My name is Eric Williams. I'm a 22 Senior Vice President at TerraPower, and I'm the 23 design authority for the Natrium project. And I'm 24 going to be talking today about a plant design 25 overview with three main points really to provide an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 1 overview of the plant design and operation, including 2 both the nuclear island and the energy island. We're 3 going to be focusing on more of the innovative 4 features that we have, you know, comparing us to a 5 light water reactor or something operating today, what 6 would be the key differences for us, and also some 7 differences between historical sodium fast reactors 8 and the Natrium reactor, as well. And then I'll hand 9 it off to Ryan, and he'll talk about the licensing 10 strategy.
11 So this first slide to begin with, I'm 12 going to spend a little bit of time on this one.
13 We've talked about this one in all of our NRC 14 engagements because it really has the key safety 15 features all contained on one slide.
16 And so when you look at the Natrium 17 reactor, it is a 840 megawatt thermal core output, and 18 the key innovative features are really in that very 19 first bullet. It's a pool type reactor, as opposed to 20 a loop style reactor, so you have a large reactor 21 vessel with a large liquid inventory of sodium within 22 it. There are free surfaces within this reactor 23 vessel, so it's a little different way of thinking 24 from a loop reactor. It also uses metal fuel, a type 25 of fuel that was undergoing heavy amounts of research NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 1 and development in the U.S. legacy SFR program but 2 never got turned into a full metal core until now. So 3 there's various reactors out there that have used 4 oxide fuel. I think even PRISM was intended to, you 5 know, start out with some oxide fuel and convert to 6 metal fuel. Natrium is going to start out with all 7 metal fuel based on the research and development done 8 at EBR-II and FFTF on this metal fuel type.
9 And then it's got a molten salt energy 10 island, so that's definitely a new piece that enables 11 us to store thermal energy in the energy island coming 12 off the nuclear island and use that to generate 13 electricity at a rate demanded by the grid, as Tara 14 mentioned in her introduction. And it also provides 15 another source of large liquid inventory that also 16 adds, along with the liquid in the reactor vessel, 17 sort of a buffer to any, you know, immediate 18 challenges to core heat removal.
19 So due to those things, we'll get into 20 that a little bit more as we go through these slides, 21 and you'll see the effects of those major features.
22 But those are the three big ones.
23 And a couple of other things about the 24 metal fuel to keep in mind, it has a high degree of 25 compatibility since it's, you know, metal fuel and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 1 metal coolant. And so through a lot of the tests, the 2 amazing tests really that were done in the EBR-II 3 reactor, they demonstrated that compatibility with the 4 coolant, and so that benefits us a great deal in the 5 safety area. And then it also behaves differently, 6 you know, with regard to reactivity, feedback effects, 7 and things like that.
8 We've also done something, we've removed 9 a big source of sodium-water interaction. So a lot of 10 historical SFRs have delivered heat from the sodium 11 system directly into a steam generator to go convert 12 that to steam and make electricity. And one of the 13 major challenges to safety with that was this 14 energetic reaction between liquid sodium and water.
15 And so by imposing the energy island with a molten 16 salt system, we've taken away that safety challenge 17 between sodium and water, so that was a big 18 improvement made just in the architecture of Natrium.
19 And I've also already referred to the 20 large thermal inertia that kind of gives us that 21 simplified response to abnormal events.
22 So if we look over to the right --
23 MR. BLEY: Dennis Bley. I think you 24 mentioned it earlier, but the salt and the sodium are 25 pretty compatible? They don't have any problems when NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 1 they interact?
2 MR. WILLIAMS: They do, they do. I was 3 going to talk about that a little bit when we get to 4 the salt chemistry slide, but I'll go ahead and --
5 MR. BLEY: That's fine, that's fine. You 6 can wait.
7 MR. WILLIAMS: All right. I'll mention it 8 then. So if we look at the right side of this slide, 9 we've just simplified this to the three fundamental 10 safety functions of control, cool, and contain, and 11 I'll kind of point out the major design features under 12 each one of these. You can kind of see they follow a 13 pattern of having a normal means of performing a 14 safety function, a passive means of the safety 15 function, plus an inherent means of providing the 16 safety function.
17 And so for control, normal control, I 18 think, is quite similar to what other reactors do. We 19 have motor-driven control rods that can control 20 reactivity and control power level during normal 21 operation and during maneuvers. We also have a 22 gravity driven control rod scram similar to other 23 reactors. But when it gets to inherent reactivity 24 control, it gets different because of the design of 25 the metal fuel and the core restraint system, how the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 1 core is mechanically designed, there is a large amount 2 of inherent negative reactivity feedback fundamentally 3 in the core design, and so what that means is that we 4 can look at these events that we call unprotected 5 transients where, essentially, we postulate that none 6 of the control rods go in by gravity and, in a certain 7 number of anticipated operational occurrence type 8 events, with none of those control rods going in, we 9 can bring the reactor not down to full shutdown but it 10 can sort of self-correct itself down to a low power 11 level and stabilize just using the inherent reactivity 12 feedback. So that is an effect that was demonstrated 13 a lot in EBR-II transients and it's a factor in this 14 core design, as well.
15 MR. BLEY: In terms of, you mentioned 16 about 800 thermal --
17 MR. WILLIAMS: 840.
18 MR. BLEY: -- megawatts. Yes. So from a 19 size standpoint, I am assuming that your team has 20 optimized, this is a lot larger core and power than 21 EBR-II, so you've optimized, I think Levesque is going 22 to explore this in the closed session a little more, 23 it's about as large as you can go, let me say it this 24 way, and still get the leakage you need so that you 25 get that reactivity feedback that you're looking for.
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24 1 MR. WILLIAMS: I would say, we can get 2 into more specifics in the closed session, but I would 3 say it's not as large as you can go, but it's as large 4 as you can go with current technology without having 5 to take some leaps forward in other areas of the 6 technology, which we do plan to do. So that's a good 7 point.
8 MEMBER MARCH-LEUBA: So what's the 9 enrichment on cycle length?
10 MR. WILLIAMS: Enrichment is HALEU, so 11 we're up to, you know, 19 percent, just under 20.
12 MEMBER MARCH-LEUBA: And the cycle length 13 that you expect?
14 MR. WILLIAMS: So the initial core loads 15 are going to be one-year cycle lengths, and the reason 16 for that is because our initial fuel type that we're 17 starting the reactor with is something that we want to 18 advance using the lead test assembly program. And we 19 want to collect data, so we want to be taking fuel 20 normally at normal intervals out of the reactor and 21 sending it out for testing to form the qualification.
22 MEMBER MARCH-LEUBA: So testing on the 23 full plan power.
24 MR. WILLIAMS: Exactly. So it will take 25 several cycles of testing with the start-up fuel form NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 1 to then collect enough data to qualify and go to the 2 advanced fuel. And then cycles will expand to two 3 years and a lot less number of assemblies needing four 4 reloads.
5 MEMBER MARCH-LEUBA: And what's your 6 break-in temperature?
7 MR. WILLIAMS: It's a 510-degree C outlet 8 and 360-degree C inlet.
9 MEMBER PETTI: So the startup fuel is 10 uranium-based, but the advanced fuel is the uranium 11 plutonium?
12 MR. WILLIAMS: No, it's all uranium-based.
13 MEMBER PETTI: It's all uranium-based.
14 MR. WILLIAMS: Yes, all uranium-based.
15 Yes, yes. In the closed session, I'll talk about the 16 differences there. Any other questions on control?
17 Okay. So when we look at cooling, again, 18 well, the main thing is to hold on to the coolant, so 19 we don't have any penetrations through the reactor 20 vessel. The only penetrations into the reactor vessel 21 go through the reactor vessel head, so there's really, 22 you know, low probability of causing a loss of coolant 23 accident. I'll also talk a little bit on another 24 slide about how the pressures of the systems are 25 designed to force leaks inward towards the reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
26 1 vessel, another feature to prevent losses of 2 inventory.
3 And then we have, you know, in addition to 4 the primary heat transport system which, you know, 5 delivers heat throughout the reactor vessel, it's an 6 integrated primary system, so all of the components of 7 the primary system are inside this reactor vessel in 8 the most compact way that we can lay them out. We 9 also have an intermediate system to this that's also 10 a sodium system, so most historical SFRs do have an 11 intermediate sodium system. We do, as well. And this 12 system can remove heat through both, can remove heat 13 forced circulation or through natural circulation.
14 And it has within it a sodium-to-air heat exchanger, 15 so it's a very effective way of removing heat when 16 we're down in the refueling temperatures or modes like 17 that where we can remove heat and totally decouple 18 ourselves from the energy island entirely by isolating 19 ourselves, in fact, and remove heat during refueling 20 through this sodium-to-air heat exchanger.
21 And then in emergency situations where 22 power isn't available --
23 MR. BLEY: I'm sorry. This is Dennis Bley 24 again. Where is that thing located? Is it inside 25 this containment picture you got here?
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27 1 MR. WILLIAMS: Yes. Let me get to the 2 next slide because I'll show you real clearly where 3 that heat exchanger is located, yes. Just trying to 4 introduce the functions here and then we'll see where 5 the equipment is actually physically laid out as we go 6 through the rest of the presentation.
7 And so, finally, that's a non-safety way 8 of removing heat. When we go down to the safety means 9 of decay heat removal during a design basis accident, 10 we have something called the reactor air cooling 11 system. This is a system that goes back to PRISM 12 design, traces its lineage back there, and its way of 13 removing heat through the outside wall of the vessel 14 using natural draft air cooling. And so it's a system 15 that's pretty robust because it's always on, so we're 16 allowing a certain amount of heat to be lost through 17 the system during normal operation to get the benefit 18 of having this system always on and ready to remove 19 heat in an accident scenario. And, in fact, because 20 it's based on radiation heat transfer and, you know, 21 as a function of temperature to the fourth power, it 22 really doesn't start to kick in until you get a large 23 enough temperature difference going. And so we're 24 able to design this system to have pretty low 25 parasitic heat losses during normal operation and then NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 1 really crank in when a design basis accident happens.
2 And so that's our really robust way of removing heat 3 during design basis accidents.
4 And then, finally, on containment, one of 5 the key features here is having low-pressure systems 6 throughout the plant. And so there's really a low 7 driving force for leakage going out the reactor 8 vessel. And, of course, design basis accidents have 9 an intact primary system, so we're not dealing with 10 LOCAs and things like that, so it's really a matter of 11 keeping leakage rates as low as we possibly can 12 through the reactor vessel head.
13 One of the key features here with sodium 14 is it's affinity for radionuclides. So the fission 15 products release through failed fuel will have to flow 16 through a large depth of sub-cooled sodium, which will 17 remove the iodine and cesium and other things that 18 will add to our mechanistic source term analysis. And 19 then we have multiple radionuclide retention barriers, 20 so you've got the ability of the fuel matrix itself to 21 hold on to fission products, we've got the intact 22 primary system, we've got the cover gas above the 23 liquid sodium, and we've got several barriers outside 24 of that, physical barriers to prevent, reduce 25 releases, as well.
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29 1 MEMBER PETTI: So you didn't specifically 2 use the words functional containment as the SECY.
3 MR. WILLIAMS: Yes.
4 MEMBER PETTI: Do you think of it as a 5 functional containment?
6 MR. WILLIAMS: Yes, yes.
7 MEMBER PETTI: Okay. And there's no 8 additional containment building.
9 MR. WILLIAMS: Right, right.
10 MEMBER PETTI: Okay.
11 MR. WILLIAMS: It's formed by multiple 12 structure systems and components, yes.
13 Okay. So if you put all of that together, 14 that's where you get to a pretty simplified response 15 to abnormal events. They really all follow the same 16 pattern. There is a scram set point that gets 17 triggers and a reliable reactor shutdown takes place, 18 and then there's a smooth transition to natural 19 circulation cooling. The reactor air cooling system 20 kicks in. It takes a long time, actually, to heat up 21 all of this liquid and start to kick in the reactor 22 air cooling system, and that starts to remove decay 23 heat, essentially, indefinitely because it's just 24 using air. There's no need to replenish the ultimate 25 heat sink or anything like that. It's a low-pressure NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
30 1 functional containment that will continue to prevent 2 uncontrolled releases of radionuclides, and there's 3 really no reliance on the energy island for any safety 4 functions. There's actually no reactor protection 5 system parameters on the energy island, so there's 6 nothing out on that side of the plant beyond the 7 nuclear island that could threaten core cooling, so no 8 scrams needed from that standpoint.
9 MEMBER MARCH-LEUBA: What is the red 10 stuff, what's the green stuff, what's the blue stuff?
11 MR. WILLIAMS: Oh, yes, I do have another 12 slide, I keep saying this, that will talk about that.
13 MEMBER MARCH-LEUBA: I don't believe you.
14 (Laughter.)
15 MR. WILLIAMS: You don't believe me?
16 Okay.
17 MEMBER MARCH-LEUBA: But you keep talking 18 about this air cooling, and I'm thinking it's inside 19 of those green things.
20 MR. WILLIAMS: Okay. Yes, let me point 21 out some of this now. The next slide I have talks 22 about the flow path within the reactor vessel, so it 23 will just go a little bit more detail. Okay. So we 24 have, the reactor vessel is laid out, and it actually 25 has multiple --
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31 1 MEMBER MARCH-LEUBA: The only way to do it 2 is with a mouse, yes.
3 MR. WILLIAMS: Yes, yes, look at that.
4 All right. Okay. So there's actually physical 5 barriers that separate the reactor vessel into two 6 main pools that we call the hot pool or the cold pool.
7 The red region is meant to indicate the hot pool, so 8 that's what you see here from the top surface of the 9 reactor vessel, it goes all the way down through this 10 central region that we call the upper internal 11 spectrum because it --
12 MEMBER MARCH-LEUBA: Is that like a core 13 barrel in the --
14 MR. WILLIAMS: Yes, it's within the core 15 and the core barrel and it's bounded by, even down 16 here, by the core grid plate at the bottom of the 17 core. And then that separates you from the blue 18 region, which we call the cold pool, which goes from 19 the intermediate heat exchanger exit all the way down 20 to the reactor vessel. It's where the pumps take 21 suction from, so the pumps will take suction from the 22 cold pool and deliver it to the bottom of the core.
23 We call that the high-pressure plenum at the bottom of 24 the core, and then flow goes out, that little darker 25 shaded red box is really the heated length region of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 1 the core. You can see how small that is relative to 2 the overall height of sodium above it.
3 MEMBER MARCH-LEUBA: Give me a reference.
4 How tall is that? Two meters, five meters?
5 MR. WILLIAMS: It's about 1.3 meters, and 6 the overall height of the fuel assemblies are, like, 7 15 feet. I'm mixing units there. So flow comes up 8 through the core and into this hot pool, and then it 9 enters these heat exchangers. There's two of them.
10 They're kidney-shaped heat exchangers you see on here, 11 and the hot pool fluid is taken into these heat 12 exchangers and flows down and discharges into the cold 13 pool that's in blue. The pumps, like I said, take 14 suction from there and return the flow to the core 15 inlet.
16 So there's also free surfaces in the 17 reactor vessel, the pool reactor. So there's a liner 18 region that goes along the outside of the reactor 19 vessel that is connected to the cold pool, so that 20 means the reactor vessel wall is seeing cold pool 21 temperatures and not hot pool temperatures, and it 22 means that there's a level difference there that is 23 equal to the unrecoverable pressure drop through the 24 IHX. So there's a level difference there, and when 25 you shut the pumps off that equalizes, yes.
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33 1 So those are kind of some of the unique 2 things about the way the reactor vessel is laid out.
3 And in some of our analyses, we also call something a 4 warm pool, but I won't get into that. That's just 5 sort of an intermediate region that goes around the 6 IHX.
7 VICE CHAIR KIRCHNER: The intermediate 8 coolant loop is a salt loop, right?
10 VICE CHAIR KIRCHNER: So you've made the 11 statement that the nuclear island's --
12 MR. BLEY: Use your mike, please, Walt.
13 VICE CHAIR KIRCHNER: Sorry. It's on 14 Dennis. I'll get a little closer to it. That the 15 nuclear island is separated from -- what do you call 16 the balance of the plant here?
17 MR. WILLIAMS: Energy island.
18 VICE CHAIR KIRCHNER: Energy island.
19 MR. WILLIAMS: Yes.
20 VICE CHAIR KIRCHNER: Do you have to worry 21 about freezing your sodium, or is the salt melting 22 point in the energy island at a high enough 23 temperature that you don't have to worry about 24 overcooling transients? That's a way that that non-25 nuclear island could couple in an adverse manner with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
34 1 your nuclear island.
2 (Whereupon, the above-entitled matter went 3 off the record at 1:38 p.m. and then went back on the 4 record at 1:38 p.m.)
5 MR. WILLIAMS: Okay. All right. So there 6 was, I think there was kind of two questions in there.
7 One had to do with are you worried about freezing, and 8 the other one about overcooling. And I think we are, 9 of course, looking at overcooling as part of the 10 safety analysis. You always have to do that, so, you 11 know, pump overspeeds and heat exchanger being more 12 effective than you intend them to be or turning on, 13 blowers turning on when you don't expect them to are 14 all accident initiators that we've looked at and don't 15 see anything real problematic there, again, because of 16 all that thermal inertia that we have.
17 Freezing is certainly something that we 18 have to manage and engineer our way through. During 19 normal operation, there's not a challenge to freezing 20 because everything is nice and hot. But if you were 21 shut down for an extended period of time, we have heat 22 tracing on all of the sodium and salt piping to 23 prevent freezing from happening, and it's also there 24 to allow for maintenance because, when you want to do 25 maintenance, you do want to freeze to actually allow NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 1 you to perform maintenance. So there's a lot of heat 2 tracing in the reactor for that reason. It's pretty 3 simple but --
4 MEMBER HALNON: The energy island, though, 5 if it freezes up, it's just an economic issue, it's 6 not a safety issue, correct?
7 MR. WILLIAMS: Right. It's not because we 8 would run back and decouple, close the isolation 9 valves entirely.
10 MEMBER HALNON: Yes, the inherent self 11 island, self nuclear island.
12 MR. WILLIAMS: Right. Okay. So there was 13 a question on ancestries --
14 MR. BLEY: This is Dennis Bley. I don't 15 know if you guys know it, but the meeting timed out 16 about five minutes ago and we can only listen on the 17 phone now. We can't --
18 CHAIR REMPE: Dennis, Dennis, most people 19 have rejoined. There was something that just happened 20 with the software and we got all kicked off for 21 something, but, anyway, most people have been able to 22 rejoin. And I think the court reporter, the court 23 reporter, just to verify, you can hear, right? Right.
24 Thank you.
25 MR. WILLIAMS: Okay. Yes, so if you look NEAL 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 1 back at all the sodium fast reactors that form a basis 2 for Natrium, you can go all the way back to EBR-I, 3 although that was really just a demonstration of 4 breeding and wasn't really sodium cooled, it was 5 sodium potassium cooled. But EBR-II and FFTF are the 6 key ones for us. EBR-II with a lot of the passive and 7 inherent safety tests, the run beyond cladding breach 8 tests. And FFTF certainly in terms of all the 9 operating experience with sodium systems and also the 10 tests that they were performing on metal fuel and 11 scram to natural circulation tests. And then even 12 TREAT, the transient reactor test facility, is a 13 source of information on severe accidents with metal 14 fuel.
15 So those are the main areas to look at for 16 the ancestry for Natrium. And like was said earlier, 17 it's a combination of the PRISM and the traveling wave 18 reactors, and I also have to mention Clinch River 19 because the Clinch River breeder reactor had certain 20 parts of its design that are beneficial to us, as 21 well.
22 All right. We spent a lot of time on that 23 slide, but there's a lot on there, so I think that was 24 good. So I think I've talked about most of this, you 25 know. Key differences from light water reactors, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 1 we've kind of already discussed that. The systems are 2 very compact. You'll see that on the next few slides 3 visually. The low-pressure systems, the efficient 4 heat transfer just because you've got metal fuel, 5 metal coolant, a metal reactor vessel. It's metal, 6 metal, metal, so it's very efficient if you draw the 7 thermal resistance diagram. Pool design with a large 8 coolant inventory is a big factor in safety.
9 Modularity, we've used modular construction wherever 10 we can. It's not a factory-built reactor vessel, but 11 we still use modularity in a lot of the design of the 12 plant and it's architected to really separate safety 13 and non-safety with special treatment systems away 14 from non-safety systems.
15 And also that energy island, nuclear 16 island separation allows for parallel construction 17 techniques, and the extra inherent and passive safety 18 allows us to do a reduced emergency planning zone.
19 MEMBER PETTI: So is it a goal for the EPZ 20 to be the site boundary?
21 MR. WILLIAMS: Yes. All right. So this 22 slide gives you a good view visually of the layout.
23 This is pretty true to what this area in Wyoming 24 actually looks like, so pretty nice view of the plant.
25 And you can see in the center of the slide, I'll use NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
38 1 my mouse here again, we have the reactor, the three 2 main nuclear island buildings. There's the reactor 3 building in the middle, there's the reactor auxiliary 4 building just next to it, and the fuel handling 5 building just next to that.
6 So those three buildings form the central 7 point of the nuclear island, and the heat essentially 8 comes from below grade in the reactor building out of 9 the reactor vessel through those intermediate heat 10 exchangers that we saw in the other slide, enters the 11 intermediate heat transport system, which is mostly 12 located in the reactor auxiliary building, so the 13 intermediate sodium goes through these sodium to salt 14 heat exchangers and delivers the heat to the salt 15 piping, which leaves the reactor auxiliary building 16 where my mouse is going all the way over to the 17 thermal energy storage tanks, which are over here on 18 the energy island. The thermal energy storage tanks, 19 well, the hot tank essentially pumps salt into the 20 salt-to-water steam generators that are in the steam 21 generator building. Those will generate super heated 22 steam to go to the turbine.
23 And everything that you see beyond the 24 energy storage tanks are sized for 500 megawatts 25 electric, even though the nuclear island is generating NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 1 345 megawatts electric steady state. So that allows 2 us to ramp up to 500 megawatts electric for about five 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if the grid demands it, but it also allows us to 4 go down to about 100 megawatts electric if the grid 5 demands it and the renewables are working the way 6 they're supposed to and the sun is shining and the 7 wind is blowing, you may want to reduce that heat and 8 save it up for later. So that's the magic of the 9 energy island and the ability to load follow with that 10 while keeping the nuclear island constant.
11 MEMBER HALNON: A couple of things jump 12 out. First of all, you may have your reasons, but I 13 would put the control building more centralized on the 14 other side of the plant just so you can get to energy 15 island quicker. But, nevertheless, if you had a two 16 unit or three or four multiple unit site, what 17 synergies would you have with this single unit besides 18 the switch yard?
19 MR. WILLIAMS: Yes, exactly. So the dual 20 unit site would essentially share the fuel handling 21 building, and so, in fact, one of our requirements is 22 to design the fuel handling building so that a later 23 second unit could be added at a later time.
24 MEMBER HALNON: Just on the other side.
25 MR. WILLIAMS: Just on the other side.
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40 1 Exactly, yes. And they can also share the energy 2 island. Our site is Wyoming is actually sized to be 3 able to add an additional pair of energy storage 4 tanks.
5 So that's the other thing you can do is 6 the energy island is very modular. There's a lot of 7 optionality on the energy island, depending on the 8 grid and where you're locating the plant, to have 9 multiple salt tank pairs, have a different sized 10 turbine, have multiple turbines. There's lots of 11 things you can do with that to meet whatever that grid 12 demands.
13 MEMBER HALNON: So just back to my first 14 comment, is there a reason why the control building is 15 not centralized? Just as an ex-operator, I would want 16 to be able to get to the energy island rather quickly.
17 MR. WILLIAMS: I think that's a good 18 comment. I mean, I know there were a lot of factors 19 into placing that. I don't recall exactly. I mean, 20 it was safety of providing the cabling that goes from 21 the control building into the reactor building. It 22 also houses the reactor protection system equipment in 23 there, so I know there were a lot of factors in that, 24 but I don't recall exactly why it's there.
25 MR. WERNER: I could maybe help a little NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 1 bit. This is Mark Werner from TerraPower. One thing 2 to note is that the operators in our control room 3 don't have any safety-related actions, so they don't 4 need to be close to anything to do anything. We do 5 have control, something like a control room in the 6 energy island to provide local control of, you know, 7 near the turbine facility, so there will be people in 8 that area that will provide local control. We are a 9 dispatchable power type of plant, so there's a little 10 bit less to do over there.
11 But one of the things we did try to do on 12 this plant, as Eric mentioned, is a little bit more of 13 a distributed architecture. From a construction 14 standpoint, that control room can be constructed with 15 a work front at the same time as the other facilities.
16 And we see a lot of value in shortening our overall 17 construction duration from a delivery standpoint. So 18 by separating all these buildings, providing a little 19 bit of space, we can have four or five separate crews 20 working on facilities at the same time. If you put 21 them all on top of each other, then you have to wait 22 for the mechanical team to finish, the civil team to 23 finish, before you can start laying down your 24 electrical work. By separating them, they can be 25 working at the same time, shortening construction.
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42 1 MEMBER HALNON: Thanks. Acreage-wise, 2 what is that about, what we're looking at? Five, five 3 acres?
4 MR. WERNER: The overall site is about 50 5 acres. The nuclear island itself is about 16 acres, 6 and it's still fairly distributed, you know. Eric 7 mentioned kind of the flexibility of our layout and 8 flexibility of our energy island, you know. How we 9 orient all of these facilities is really not coupled 10 to our safety case at all, and so if we had to deploy 11 a site not like Wyoming and space was a real concern, 12 we could make things closer.
13 VICE CHAIR KIRCHNER: You want to orient 14 your turbine building in a preferential way.
15 MR. WERNER: Sure. Right. And to be 16 clear, there are spacing requirements and location 17 requirements. I'm just noting that, you know, if we 18 go to a site that was narrower or longer, because 19 we're decoupled and that salt pipe doesn't lose a lot 20 of heat, isn't that expensive in the grand scheme of 21 things, we can reorient our two sites to kind of fit 22 the local geography pretty well.
23 VICE CHAIR KIRCHNER: Dennis, I see your 24 hand up. Go ahead.
25 MR. BLEY: Yes, I had a couple of things.
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43 1 One is I kind of agree with Greg. I don't know how 2 long the distance is from the control building over to 3 the energy island, but having a lot of separation from 4 the operators could be troublesome at some times. I 5 understand your construction ideas, too.
6 The idea is, I really like this. This is 7 pretty clever. The other side of the story, if the 8 grid load drops off or the grid trips, it would appear 9 you don't need to scram, you can keep pumping heat 10 over into those energy storage tanks. How long can 11 you do that before you got to ramp down power or trip 12 the reactor?
13 MR. WILLIAMS: Yes, it's essentially 14 another, it's just five hours in both directions.
15 MR. BLEY: Either way. Okay.
16 MR. WILLIAMS: Yes.
17 MR. BLEY: And you run into some kind of 18 mechanical limits on the temperature if you --
19 MR. WILLIAMS: Right. We would be looking 20 at the tank levels, and the tank levels would get to 21 a minimum or maximum --
22 MR. BLEY: Sure.
23 MR. WILLIAMS: Yes.
24 MR. BLEY: Okay. Oh, by the way, please 25 keep talking into your microphones. When you turn NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 1 your head away, we can't hear you out here.
2 MR. WILLIAMS: Oh, okay. Thanks.
3 MEMBER MARCH-LEUBA: We're having problems 4 today that we typically don't have. On the storage 5 tanks, do you increase the temperature of the salt or 6 do you melt more salt? You said something about 7 level.
8 MR. WILLIAMS: Yes, we don't melt more 9 salt. The salt tank levels change, so, if you are 10 ramping down electrical output, you accumulate hot 11 salt in the hot salt tank, so the level goes up.
12 MEMBER MARCH-LEUBA: So it's just 13 expansion.
14 MR. WILLIAMS: Yes.
15 MEMBER MARCH-LEUBA: So you're changing 16 the temperature of the salt, not the amount of salt.
17 MR. WILLIAMS: It's the amount of salt in 18 that hot salt tank, yes. It will go up. You will 19 accumulate hot salt if you're throttling back the hot 20 salt --
21 MEMBER MARCH-LEUBA: So what is it the 22 cold salt?
23 MR. WILLIAMS: And the cold salt is in the 24 cold salt tank --
25 MEMBER MARCH-LEUBA: Oh, so you have two NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
45 1 tanks, one is cold, one is hot.
2 MR. WILLIAMS: Yes.
3 MEMBER MARCH-LEUBA: So you are, it's not 4 that you're melting salt.
5 MR. WILLIAMS: Right.
6 MEMBER MARCH-LEUBA: You're transferring.
7 MR. WILLIAMS: Moving it, yes, yes. It's 8 a constant volume of salt.
9 MR. LEVESQUE: It just, you know, it 10 transfers from one tank to the other as we turn hot 11 salt into cold salt.
12 MR. WILLIAMS: Dr. Rempe, what do we do 13 when we get to 2:00? Do you want me to keep going?
14 VICE CHAIR KIRCHNER: We'll stop where we 15 are, and then we will have people who are joining us 16 and are okay to join us will come in on a different 17 Teams connection --
18 MR. WILLIAMS: Okay.
19 VICE CHAIR KIRCHNER: -- that's been 20 limited. So we'll need to take a break, and we'll 21 need to just make sure everyone in the room, I see 22 you've handed out already proprietary information. I 23 hope you've got a handle on it because the room has 24 been open, but we'll close the room, and then we'll go 25 into the closed session.
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46 1 So if you're at that juncture, this would 2 probably be a good time for us to take a 10- to 15-3 minute break and reset our Teams connection.
4 MEMBER MARCH-LEUBA: It doesn't need to be 5 exactly at 2:00. It can be at 2:30.
6 CHAIR REMPE: I hadn't heard it had to be 7 exactly.
8 MR. MOORE: This is Scott Moore, the 9 executive director. It does not have to be precisely 10 at that time. The meeting in Teams will remain open.
11 VICE CHAIR KIRCHNER: Okay.
12 CHAIR REMPE: But when you want to go to 13 the closed meeting, it doesn't happen quickly because 14 we've got to check the room and a break is good idea.
15 MR. WILLIAMS: Okay. All right. So this 16 was the vertical cut-through, those central three 17 nuclear island buildings, and so right away you can 18 kind of see that we have the, like I said, the safety 19 related and non-safety with special treatment systems 20 below grade. All the radiological systems are 21 essentially below grade in the reactor. That includes 22 the reactor vessel, it includes the piping of the 23 intermediate heat transport system just above the 24 reactor vessel. And this area that you see, this 25 rectangle above the reactor vessel is what we call the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 1 head access area, so there's a lot of equipment, very 2 important equipment in that region.
3 Over below grade in the reactor auxiliary 4 building, we've got radiological systems there for 5 processing both the liquid sodium from the reactor 6 vessel and the gaseous, the gases from the reactor 7 vessel, so those are all below grade. And then above 8 grade, you've got the sodium-to-salt heat exchangers, 9 essentially, and the intermediate sodium pumps that 10 make up the intermediate heat transport system.
11 You can also see on the right the fuel 12 handling building, spent fuel pool below grade. And 13 then in both the reactor building and fuel handling 14 buildings, you can kind of see they're very open 15 structures, so the refueling maneuvers can happen 16 between these two buildings. It's not that we've 17 forgotten to add systems in these buildings. They 18 really are open. And you can see that, because the 19 radiological systems are below grade in this nuclear 20 seismic grade concrete structures, we can afford to 21 take a different approach above grade, and we have, 22 essentially, steel structures above grade. So that's 23 where a lot of the modularity comes in, too. These 24 steel structures, you know, can be handled much 25 differently and economically.
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48 1 I had a question earlier about where was 2 the intermediate air cooling heat exchanger, and this 3 was the slide that shows that very clearly. Over here 4 on the side of the reactor auxiliary building, outside 5 of the reactor auxiliary building, and so that is a 6 sodium-to-air heat exchanger connected to the 7 intermediate heat transport loop and it allows through 8 both forced cooling and natural draft cooling to take 9 heat out of the reactor vessel and deliver it to the 10 air without using the safety related system. And so 11 that's the normal decay heat removal system that we 12 would use in refueling modes and hot standby and so 13 forth.
14 You can also see the ducts for the reactor 15 air cooling system here, and so the air is flowing 16 down and around the reactor vessel and it's coming in 17 through ducts, and these lower ducts is where it's 18 coming in and it's coming out through the taller 19 ducts. And there's four inlets and four outlets to 20 that system.
21 VICE CHAIR KIRCHNER: What parts of the 22 reactor building do you have to harden against natural 23 hazards in particular?
24 MR. WILLIAMS: Yes, that would be this 25 level at grade that protects the --
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49 1 VICE CHAIR KIRCHNER: So your reactor air-2 cooling ducts, if they were severed for whatever 3 reason, the system would still function?
4 MR. WILLIAMS: Right. We are looking at 5 that, and they're set up on different sides of the 6 reactor building with enough separation that we can 7 deal with some of those hazards.
8 VICE CHAIR KIRCHNER: One of the 9 vulnerabilities that was, I think, there in the PRISM 10 design was that they had that functional heat 11 exchanger above grade, potentially exposed to hazards.
12 So, basically, if these ducts sever, you'll still get 13 a natural draft through the reactor cavity to take out 14 the decay heat.
15 MR. WILLIAMS: They are very robust with 16 regards to, you know, if you sever two of them and you 17 still have two functioning or if you increase pressure 18 drop in these ducts. A lot of the testing that was 19 done on the system was done at Argonne National Lab 20 and showed that it really wasn't as sensitive to 21 pressure drop as you would think. It's more radiation 22 heat transfer limited, and so it's a very unique kind 23 of robust system. But there's all kinds of external 24 hazards also that can degrade the system that we look 25 at, you know, debris getting entrained into the ducts, NEAL 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 1 and there's all sorts of degradation mechanisms that 2 were taken into account to --
3 VICE CHAIR KIRCHNER: Dennis, do you have 4 your hand up again?
5 MR. BLEY: I do. This is Dennis Bley.
6 VICE CHAIR KIRCHNER: Okay. Go ahead.
7 MR. BLEY: It looks like if you've 8 hardened the aux building, things are probably pretty 9 good, except airplane crash, I assume, could take out 10 the intermediate air cooling system; is that right?
11 MR. WILLIAMS: The intermediate, there's 12 two intermediate air cooling heat exchangers on sides 13 of the reactor aux building, so I'm not, I can't 14 answer that exactly whether it can take out both of 15 them or not; I'd have to go back and look. But those 16 aren't safety-related systems either.
17 MEMBER MARCH-LEUBA: Your safety-related 18 air cooling is inside of the reactor cavity.
19 MR. WILLIAMS: Right.
20 MEMBER MARCH-LEUBA: And where does the 21 heat go?
22 MR. WILLIAMS: So the heat --
23 MR. BLEY: Yes, you were going to show us 24 that.
25 MR. WILLIAMS: Yes. So the heat, it comes NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 1 in through the reactor air-cooling ducts that I'm 2 pointing to with my mouse. There's two of them shown 3 here. There would be two, a mirror image, on the 4 other side. So four air intake ducts.
5 MEMBER MARCH-LEUBA: That reactor cavity 6 has a couple of holes that are connected with the 7 atmosphere.
8 MR. WILLIAMS: Absolutely. It's all 9 connected to the atmosphere. And so it comes down --
10 MEMBER MARCH-LEUBA: The one, it looks 11 like a containment fluoride. It's not.
12 MR. WILLIAMS: Right. Well, yes. So let 13 me try to --
14 MEMBER MARCH-LEUBA: You have a way for --
15 I couldn't figure out where the heat was going.
16 MR. WILLIAMS: Okay. We do have more 17 sketches of this in the closed session that show the 18 details.
19 MEMBER PETTI: Do you use a seismic 20 isolator on the --
21 MR. WILLIAMS: We don't seismically 22 isolate the plant at large, but we have seismic 23 isolation on the reactor vessel.
24 MEMBER PETTI: The reactor vessel. Good.
25 MR. BLEY: I guess, before you leave this, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 1 I'm still thinking about an aircraft crash, and that 2 reactor building up there, that's a steel building it 3 looks like. If that gets taken out, it seems like 4 those air ducts and things could all get crushed up 5 and wouldn't work anymore.
6 MR. WILLIAMS: Yes, we'll have to show 7 that there's enough air cooling even with an airplane 8 crash.
9 VICE CHAIR KIRCHNER: But you wouldn't 10 want collateral damage if you have an aircraft event 11 hitting that sodium intermediate loop and then 12 spreading sodium into the reactor building cavity.
13 MEMBER MARCH-LEUBA: On a rainy day.
14 MR. WILLIAMS: Yes, on a rainy day.
15 MR. BLEY: Okay. We'll see that kind of 16 detail later, but I'm curious about how you deal with 17 that.
18 MR. WILLIAMS: Oh, Tara did mention 19 something to me real quick. There's been an evolution 20 to this design, and the reactor, it looks like the 21 reactor air-cooling ducts are integral with the 22 reactor building, but they're actually separate from 23 the building. So the reactor air-cooling ducts are 24 safety-related ducts, and the reactor building is not.
25 So those are separate.
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53 1 MR. BLEY: Okay. There's a great big 2 crane up there that could come down with the building, 3 though, it looks like. Pretty heavy.
4 MR. WILLIAMS: We'll jump into 3D model in 5 the closed session, and we can look at this in detail 6 of the current design. This is a pre-conceptual 7 image.
8 Okay. So I'll go through this one a 9 little quickly. We talked a bit about the energy 10 island thermal storage. The great part about this 11 side of the plant is that all of this technology is 12 readily available from the concentrated solar plant 13 industry, and so that was one of the benefits of why 14 we chose it. We even chose the same salt composition 15 that is typically used in concentrated solar plants to 16 make their, you know, good access to the OE database 17 of all of that work.
18 So when you look at this, there's lots of 19 different options, from the number of tanks, the 20 number of steam generator trains, and what sized 21 turbine you need for the grid where you want to put 22 this plant. The picture on this slide on the left 23 kind of shows you a top-down view of a concentrated 24 solar plant, and you can see how many salt tank pairs 25 they have. So they've definitely chosen to go with a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 1 higher degree of thermal energy storage for this 2 location, and, like we said earlier, that's something 3 that you can do with the Natrium design, as well.
4 All right. So a few things to just talk 5 about the benefits of sodium and then the benefits of 6 salt. When we look at sodium coolant, and there's 7 been a lot of work with liquid sodium and liquid 8 sodium reactors in the United States, the one that's 9 really not, that's not really mentioned on this slide 10 is that sodium is chosen for its neutronic properties 11 because it doesn't moderate neutrons. So if it were 12 to have a higher degree of moderation, it would soften 13 the spectrum, and we may not have a fast reactor 14 anymore. So it's chosen for the reasons of wanting 15 the sodium fast reactor, and then the higher heat 16 capacity that it provides allows us to have reasonable 17 sized pumps. The high heat transfer is really the 18 most important thing. It allows the core to be 19 reasonable sized, and it allows that great decay heat 20 removal that I talked about and allows us to use these 21 sodium-to-air heat exchangers that are so effective 22 and so valuable to us in decay heat removal.
23 It's also got a good range for where we 24 want to operate, so the high boiling point around 883 25 degrees C and a melting point around 98 degrees C, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 1 all of that is at atmospheric pressure, so that means 2 we can have a high temperature heat output coming from 3 the reactor vessel, which allows us to have high 4 thermal efficiency without having to pressurize it.
5 So that's the key to being at atmospheric pressure.
6 And then the density of sodium is 7 remarkably similar to water, which allows us to do a 8 lot of testing, do the scaling of a test for sodium in 9 a water system and actually get that to work, and so 10 reducing the costing of testing and increasing the 11 value of testing with water. Lack of corrosion, so 12 there's been demonstrations of very low corrosion in 13 sodium systems over time. We learned a lot of that 14 from FFTF. Limited auxiliaries, like I mentioned, a 15 couple of systems to remove non-condensable gases and 16 impurities from the sodium liquid and the gas; but 17 other than that, there's not a whole lot of 18 conditioning that you have to do.
19 And then there's a large sodium inventory, 20 so 800 cubic meters of sodium in the reactor to absorb 21 all of that decay heat removal when you need it. And 22 the graphs here show you the differences between, 23 like, the peak, the cladding temperature and the peak 24 central temperature of the fuel between an oxide fuel 25 and metallic fuel. And so because metallic fuel can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 1 remove heat so effectively from the fuel itself, you 2 actually have a pretty close delta between the fuel 3 temperature and the cladding temperature shown on the 4 right, whereas an oxide fuel you don't. You have a 5 higher center line fuel temperature that is sometimes 6 above the boiling temperature of the coolant. And in 7 sodium reactor oxide fuel reactors, they have that.
8 So the normal operating fuel temperature was higher 9 than the boiling temperature, but in Natrium it's 10 several hundred degrees below the boiling temperature 11 because of metal fuel. So those were some of the 12 great benefits of using sodium.
13 And then this next slide just shows some 14 pictures. If you come out and visit us in our lab, 15 we'll let you take a look at these up close but not 16 too close, and we even let people cut into solid 17 sodium to kind of get a sense of what it feels like.
18 It kind of feels like cutting through butter taken out 19 of the refrigerator or something. But you can see 20 it's opaque, so that's kind of an important property 21 that we have to be concerned with in terms of 22 refueling. A lot of times, it builds up an oxide 23 layer on the top. It has an auto-ignition temperature 24 that varies quite widely between about 100 degrees 25 Celsius and 400 degrees Celsius. So, you know, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
57 1 could have sodium sitting there below its auto-2 ignition temperature, but, of course, at reactor 3 temperatures, it's going to be above that.
4 And on the solid side, you can kind of see 5 it's built up this oxide film. It's kind of grayish 6 or even pinkish sometimes.
7 And so one of the things with sodium is it 8 does react energetically with both water and air, and 9 so, for our sodium systems, we have to prevent and be 10 prepared to detect and mitigate any sodium fires. So 11 one of the downsides to using sodium is, in addition 12 to the heat tracing we mentioned earlier, is providing 13 all that sodium fire protection.
14 And so, you know, it's a kind of 15 engineered safety system that has to exist in the 16 plant. But the good news is it's all at atmospheric 17 pressure, so, when we have leaks, they're not, you 18 know, you sometimes see in the literature spray fires 19 and pool fires and things like that. What we 20 typically call it is a drift fire. It would be like 21 a slow leak at a flange or something like that.
22 Typically, these leaks are, you know, they don't get 23 through the guard piping and inerted space that we 24 have the pipe in, and so they're detected by the 25 sodium leak detectors, which then allow us to go in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 1 there and perform the maintenance --
2 MEMBER PETTI: So all the pipes are 3 guarded.
4 MR. WILLIAMS: All the pipes have some 5 kind of, and we have a slide on this, some kind of 6 leak protection to them. There's all different ways 7 depending on where the pipe is.
8 MEMBER PETTI: The opacity I was aware of, 9 but the program inspection and, you know, how you see 10 into this code requirements. I know, obviously, some 11 of the components are above the level, but you 12 probably still have a need to get down lower with the 13 sodium. Has the technology evolved that you've got a 14 solution?
15 MR. WILLIAMS: We're still looking at, you 16 know, technologies for under sodium viewing, so that 17 is certainly a part of a project is to look into 18 those. But we're also looking at, you know, risk-19 informed inspections and different ways of inspecting 20 equipment that is below the level of sodium, so it's 21 kind of a combination of all of them. But for the 22 refueling, you know, we will have discrimination 23 features on the top of the fuel assemblies that will 24 allow all of that to occur, even under the opaque 25 sodium. And fuel assemblies are inspected when they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
59 1 come out and they get cleaned and all of that, as 2 well.
3 MEMBER BALLINGER: This is Ron Ballinger.
4 Are you in any way connected with or taking advantage 5 of the French side, especially with respect to 6 inspection, refueling, those kinds of things with 7 Phoenix and Super Phoenix?
8 MR. WILLIAMS: We're still working on that 9 front. We'd like to be more connected. I don't think 10 we are -- but we are talking right now.
11 MR. WERNER: I think we're sending a team 12 over, I think we're sending a team over, like, next 13 month to collaborate on methods and operations, yes.
14 And this is Mark Werner.
15 VICE CHAIR KIRCHNER: Also look at 16 construction, too, because if you study the Super 17 Phoenix experience.
18 MEMBER BALLINGER: They had some infamous 19 issues, like not being able to find the fuel elements.
20 MR. WILLIAMS: Okay. That would be 21 infamous. All right. So another couple of slides 22 just on salt. So we are using nitrate salt, you know, 23 60-percent sodium nitrate, 40-percent potassium 24 nitrate. This is the typical composition used in the 25 solar industry. Molten salt inventory is quite large NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
60 1 with those large thermal storage tanks, 30,000 tons.
2 So there's quite a bit of thermal energy storage 3 capacity that we have over there, and, of course, 4 anything happening on the other side of those tanks 5 with the steam generators, the turbine, or all of that 6 equipment, just can't make its way back to the nuclear 7 island very quickly.
8 So there's lots of benefits of using this 9 molten salt. It actually does have pretty good long 10 life with really low performance degradation, and it's 11 really good in the range of temperatures that we want.
12 We can use carbon steel up until about 400 degrees C 13 and stainless steel up to about 600 degrees C, so the 14 cold tank side is typically carbon steel and the hot 15 salt tank side is typically stainless steel. So a 16 really good temperature range for us and a really high 17 degree of efficiency in terms of retaining the heat.
18 And so a lot of people ask us, you know, are you 19 losing a lot of heat from the tanks that are sitting 20 there, you know, during outages and things like that 21 and not being used, and the answer is no. Very little 22 heat is lost through them.
23 And so, you know, a really common use in 24 the industry. I don't think it's been used in a 25 nuclear facility, so, you know, we are taking a look NEAL 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 1 at, you know, the operating experience and history 2 with designing these tanks in the industry and making 3 sure that they are designed to the right level of 4 reliability for us.
5 MEMBER BALLINGER: Do you have any 6 flammability issues with this salt mixture?
7 MR. WILLIAMS: No. So salt is not 8 flammable with air, so when salt leaks, you know, you 9 go and clean it up with a shovel. It solidifies on 10 the outside, and there's no danger with that.
11 I was going to mention the salt and the 12 sodium is an exothermic reaction. It's relatively 13 mild. It's nothing like sodium and water, but it is 14 one of the technology factors that we are doing quite 15 a bit of testing on right now in our lab in 16 Washington. And so we have a couple of systems, a 17 couple of test loops running right now with sodium-18 salt interaction tests, and what the data are really 19 showing us is that when you have, in our case you 20 would have salt leaking into a sodium system, it does 21 generate some energy, but it is relatively mild. And 22 because the sodium is so thermally conductive, the 23 heat dissipates very quickly in the sodium system, and 24 you can detect it by the presence of, I think it's 25 hydrogen and nitrogen that are generated from this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
62 1 reaction, and those are gasses that we're already 2 looking at in the intermediate heat transport system.
3 So you can detect a leak when it happens, and you can 4 shut down and perform maintenance. You can train the 5 heat exchanger and perform maintenance on that. But 6 we're still continuing to test that.
7 Yes, go ahead.
8 VICE CHAIR KIRCHNER: Go ahead, Dennis.
9 MR. BLEY: You said hydrogen. This is 10 Dennis Bley. I don't see any hydrogen.
11 MR. WILLIAMS: It's the -- oh, yes, you're 12 right. Yes, it must have been nitrogen. Sorry.
13 You're right. There's no hydrogen in there. Thank 14 you for correcting me on that.
15 MEMBER PETTI: And there's no chemical 16 issues in terms of the vapor pressure that workers 17 need protection, besides the temperature of the salt, 18 in terms of maintenance.
19 MR. WILLIAMS: Yes, I don't, I'm not aware 20 of any.
21 MEMBER BALLINGER: There was an incident 22 at Cadarache where they were cleaning a tank which 23 formerly contained sodium and had residue in it, and 24 something happened. Now, I'm a little fuzzy on what 25 they were using, whether it was a cleaning fluid or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
63 1 whatever it was, reacted with the residual sodium, and 2 they killed a few people.
3 MR. WILLIAMS: Okay.
4 MEMBER BALLINGER: It was an enclosed 5 tank, and it was a bad hair day.
6 MR. WILLIAMS: Yes.
7 MEMBER BALLINGER: So there are those 8 issues.
9 MR. WILLIAMS: Yes. I think for sodium 10 you have to take a lot more precautions than with the 11 salt.
12 MEMBER BALLINGER: Certainly with 13 cleaning, with what they were doing.
14 MR. WILLIAMS: Right, right. Good 15 comment.
16 So salt in the molten state, you know, 17 it's the white substance on the left as a solid and a 18 clear liquid as you can see on the right. And so, you 19 know, leaks are, again, something that we have to 20 consider and look at, but nothing like sodium leaks.
21 So we'll definitely be looking at that and having 22 plans for maintenance and recovery after leaks.
23 MR. WERNER: Hey, Eric, this is Mark 24 Werner, just jumping in. If you go back a slide, just 25 to maybe add a little more context to this. The image NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
64 1 on the left is actually salt that we leaked out into 2 a pan and just collected in a jar. And so when we're 3 worried about leaks, it just solidifies to look like 4 what's on the left. And on the right, you know, 5 that's molten salt, but it's in an air environment.
6 It's not in a glove box or anything, so this kind of 7 image shows how compatible this working fluid is with 8 the environment.
9 MR. WILLIAMS: Thanks, Mark. All right.
10 So when you take a look at the safety systems, we 11 called them, you know, compact and robust for the 12 Natrium reactor. So if you just look at comparison to 13 a light water reactor where you're, you know, your 14 safety systems are daisy-chaining multiple systems to 15 take the heat out of the reactor and take it to its 16 ultimate heat sink, it just generates a lot of, we 17 call it nuclear sprawl sometimes that makes a large 18 fraction of the plant become safety related as a 19 consequence. And so by having these reactor air-20 cooling ducts removing heat using air, you have a more 21 compact system that doesn't result in that same 22 sprawl. But having the integrated primary system all 23 inside the reactor vessel, we've eliminated a lot of 24 Section III pipe welds.
25 Being at atmospheric pressure we've NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 1 mentioned is a big feature here. And just that 2 unlimited supply of the ultimate heat sink, very 3 important for long-term cooling. Being fully passive, 4 having a system actually always in operation, and just 5 having a very rugged system. So there's a lot of good 6 benefits to the technology in that area.
7 All right. So I'm getting close to the 8 end here.
9 MR. BLEY: Dennis Bley again. I almost 10 forgot my question when the picture went away. How 11 hot does the sodium have to get -- I could have done 12 the calculation, I guess, but I didn't -- before this 13 transfer, heat transfer mechanism becomes real 14 effective? It's got to go up a hell of a lot, I would 15 think.
16 MR. WILLIAMS: Yes, yes, it's probably 17 going to go up, you know, 50 to 100 degrees, I would 18 guess. You know, it's always going to be ramping up 19 as temperature is going up. It just takes a long time 20 for that much inventory to raise its temperature. But 21 the initial, you know, following an accident with only 22 reactor air cooling available and nothing else, you 23 know, you would see a lot of the decay heat turning 24 into sensible heat, increasing the temperature of that 25 liquid pool. And then, you know, over the course of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
66 1 a day or more, you would start to see it start to kick 2 in more.
3 MR. BLEY: Okay. That gives me a hint 4 anyway.
5 MR. WILLIAMS: Okay. So this was the one 6 I was promising about the flow path. It shows kind of 7 what I described earlier, but I think it gives you 8 just a little bit of a clear image between the red hot 9 pool and the blue cold pool. And the arrows show you 10 the flow path through the heat exchangers and the 11 pumps.
12 So a couple of features I'll just point to 13 that we haven't talked about yet is there is a guard 14 vessel outside of the reactor vessel. So one of the 15 ways of preventing, you know, if there was a leak, and 16 this would be a beyond design basis leak from the 17 reactor vessel, we don't want to have interaction with 18 air. So there's a guard vessel surrounding the 19 reactor vessel and an inerted space in between them.
20 And then the gap between that reactor vessel and guard 21 vessel is sized such that, if there was a reactor 22 vessel leak, the total liquid level would still remain 23 above the pumps and the heat exchangers inside the 24 reactor vessel. So even if you had a reactor vessel 25 leak, you'd keep those components covered and you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
67 1 provide the opportunity to continue to use forced 2 cooling, if you have it, and to still allow natural 3 circulation cooling to occur, of course.
4 MEMBER MARCH-LEUBA: Are the pumps on the 5 outlet of the heat exchanger?
6 MR. WILLIAMS: The pumps take suction from 7 the cold pool and discharge to the core inlet.
8 MEMBER MARCH-LEUBA: Kind of push it. And 9 what drives it through the heat exchanger?
10 MR. WILLIAMS: The heat exchanger takes 11 input from the hot pool at the top and then discharges 12 it into the cold pool.
13 MEMBER MARCH-LEUBA: By gravity?
14 MR. WILLIAMS: Through the pump. The pump 15 is forced cooling up through the core --
16 MEMBER MARCH-LEUBA: So the pump is 17 cooling on it.
18 MR. WILLIAMS: Right.
19 MR. BLEY: Hey, this is Dennis Bley again.
20 On your first picture, the label on the guard vessel 21 said containment. It looks like it is essentially a 22 containment. Why don't you call it one?
23 MR. WILLIAMS: It's a major part of the 24 functional containment strategy. The guard vessel 25 surrounds completely the guard vessel, but it doesn't NEAL 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 1 go up around the reactor vessel head. And so you have 2 the seals in the reactor vessel head for the, you 3 know, for the control rod drive mechanisms, and the 4 other penetrations are also part of that functional 5 containment.
6 MR. BLEY: Okay. This picture is not a 7 real picture, so it looks like the vessel head is 8 actually bolted down on to the guard vessel. Is that 9 true, or is that just an artifact?
10 MR. WILLIAMS: Maybe Mark could help me 11 with the interface there between the guard vessel and 12 the reactor vessel.
13 MR. WERNER: Yes. This is Mark Werner 14 from TerraPower. There will be some type of seal 15 between the guard vessel and the reactor vessel. The 16 reactor head will be bolted down on to the reactor 17 support structure, but the guard vessel will be, you 18 know, sealed to the reactor vessel to maintain that 19 boundary.
20 MR. BLEY: Okay.
21 MR. WERNER: We need to keep that inert, 22 as well, and we monitor that space.
23 MR. WILLIAMS: All right. You can see, 24 the only other piece I wanted to point out here was 25 the in-vessel transfer machine shown in blue there and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
69 1 pointed out with the label. So we are doing refueling 2 inside the reactor vessel. Another kind of new thing, 3 the reactor vessel head doesn't come off. New fuel 4 comes in through an opening in the reactor vessel head 5 and gets manipulated with the in-vessel transfer 6 machine to its location; and, you know, spent fuel is 7 manipulated with the in-vessel transfer machine to 8 remove it from the core. We actually store fuel 9 assemblies after they come out of the core for a 10 period of time in in-vessel storage locations, which 11 you can think of as a couple of rings of empty fuel 12 assembly slots outside the core barrel. So they stay 13 there, they cool off for a cycle, and then the in-14 vessel transfer machine removes them out of the 15 reactor vessel. So that's a unique part of the 16 technology, and so we chose to do it with an in-vessel 17 transfer machine here for Natrium.
18 MR. BLEY: This is Dennis Bley again. In 19 the closed session, are you going to be able to show 20 us how that thing actually works? Does it index 21 around or what's going on there?
22 MR. WILLIAMS: Yes. We've got a movie for 23 you in the closed session to show that.
24 MR. BLEY: Great. I can't wait.
25 MEMBER BALLINGER: This is similar to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 1 Phoenix.
2 MR. WILLIAMS: I think this is a 3 pantograph, and I'm not sure if they use that at 4 Phoenix. But FFTF used an offset arm type of 5 equipment, and so it was different than this. And so 6 we chose to go with this because we think it will 7 allow us to have more, you know, meet the expectations 8 of commercial refueling times, as opposed to what they 9 did at a test reactor. All right.
10 VICE CHAIR KIRCHNER: So in the closed 11 session, Eric, could you show us where actually in the 12 core the spent fuel is stored temporarily?
13 MR. WILLIAMS: Yes, yes, we have that. So 14 the last slide here before I turn it over to Ryan, 15 we've kind of talked about all of this, but just to 16 wrap up the overall story for heat removal is, you 17 know, you have normal ways of removing heat, which is 18 forced flow through the intermediate heat transport 19 system, using that intermediate air heat exchanger.
20 You also have a passive way of using that same system 21 by not having forced flow and just using natural draft 22 through the intermediate heat transport system, and 23 that intermediate air-cooling heat exchanger has 24 dampers that open and a blower that can also come on 25 and force cool it if you're in the normal mode. And NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
71 1 then if you're in the passive mode, the dampers just 2 open and it allows natural draft cooling using the 3 same heat exchanger.
4 And then, finally, I call it the inherent 5 decay heat removal, and that's with the reactor air-6 cooling system because it's always on. And that's our 7 safety-related decay heat removal method.
8 So that just kind of summarizes all the 9 ways of getting heat out of the reactor, and now I 10 will turn it over to Ryan who will talk about our 11 licensing strategy.
12 MR. SPRENGEL: Good afternoon. Ryan 13 Sprengel. I'm the Director of Licensing for Natrium.
14 I did check before the meeting. It's been seven and 15 a half years since I've been in front of the 16 Committee, so I see some familiar faces and some new 17 faces. Dare I say, I'm happy to be back in front of 18 you.
19 It's been touched on already, but we are 20 using a Part 50 licensing process. So we'll go 21 through a two-cycle application with our construction 22 permit followed by our operating license application.
23 And highlighted on the slide here, in terms of the 24 approvals for the plant, you know, we'll come back 25 before the ACRS again multiple times. What was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 1 mentioned earlier, we'll touch on some of the topical 2 reports, so there's other opportunities to come in 3 front of the ACRS, as well.
4 Part of the reason for the Part 50 process 5 is tied to something that Chris mentioned. You know, 6 we're a part of the ARDP program, so there's an 7 aggressive time frame that we're working under and the 8 two-part Part 50 approach lends itself to starting 9 construction earlier and engaging with the staff 10 earlier to facilitate that time frame.
11 Currently, if we look at this slide, we 12 are to the left of it, so we're kind of off this 13 slide, you know. So we will get our construction 14 permit application submitted and start this flow path.
15 Let's see. Topical reports we'll touch on 16 in a minute, and I'll speak to those time lines 17 specifically.
18 MEMBER MARCH-LEUBA: Just a question. On 19 the construction permit, your intention is to submit 20 it before you really have a finalized design?
21 MR. SPRENGEL: Yes, yes, yes.
22 MEMBER MARCH-LEUBA: It would be kind of 23 what I call aspirational, the statements.
24 MR. SPRENGEL: I don't know if I would 25 consider aspirational, but we will be following on 10 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
73 1 CFR 50.34, speaking to a preliminary state of the 2 design. As was noted, though, our project --
3 MEMBER MARCH-LEUBA: It will mostly set 4 requirements that you have to meet on the license.
5 MR. SPRENGEL: In some areas, yes. In 6 some areas, there will be things that will be planned 7 out. Some where testing, certainly, will be ongoing.
8 The structures, after we have the CP submitted and 9 reviewed and issued as a construction permit, you 10 know, our project is set up to start construction.
11 MEMBER MARCH-LEUBA: From the typical 12 point of view, when you review a CP, you're very 13 disappointed. There is no information there. So I'll 14 focus myself to adjusting that way.
15 MR. SPRENGEL: Okay.
16 VICE CHAIR KIRCHNER: I think what Jose is 17 referring to is could you give us just maybe some 18 feeling for what the maturity of the design is when 19 you go into the TP stage? Are you ready, at that 20 point, to write procurement specs or you're still, is 21 the design evolving subsequently?
22 MR. WILLIAMS: I'd say it's definitely 23 taking the design down beyond plant functional 24 requirements to system requirements. And then the 25 safety analysis is mature enough to ensure that, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
74 1 know, the plant has enough safety margin in it to 2 start construction, so you'll see a lot of the design 3 basis accidents and even beyond design basis 4 accidents, the PRA. Because we're using licensing 5 modernization program, which Ryan is going to talk 6 about here, PRA is a very, you know, factors very 7 prevalently in the design from the beginning, so there 8 will be a lot of that information available. So it 9 doesn't mean we're done with design, but we have to 10 have established the safety margins of the design.
11 MEMBER MARCH-LEUBA: So it won't be a 12 complete PRA, but at the CP stage you expect a decent 13 PRA?
14 MR. WILLIAMS: I do, yes.
15 MEMBER MARCH-LEUBA: Okay. That's good.
16 CHAIR REMPE: You know, I was puzzled with 17 your comment, too, and I guess, you know, we've seen 18 a range of maturities in what we're seeing, and some 19 are very good and, yes, they identify what else is 20 needed, the staff does, and we review it. And then 21 they make changes, and that's wonderful. But it 22 helps, I think, to focus the questions on what you 23 might want to consider that needs to be changed.
24 That's where he's coming from, not that, you know, the 25 way you responded back was a little puzzling to me NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
75 1 that you were going to change your focus. But, 2 anyway, I think that will be good to see more detail 3 is the bottom line.
4 MR. LEVESQUE: And if I could add, this is 5 Chris Levesque, TerraPower CEO. Just a reminder about 6 how well resourced this project is, I mean, thanks to 7 the Department of Energy and the ARDP grant and thanks 8 to our shareholders. I mean, we have 800 engineers 9 working on the project today, getting into 10 considerable design detail, some of which you'll see 11 in the closed session. So there's a lot of rigor and 12 a lot of detail being established today that I think 13 you'll see in the construction permit application.
14 MEMBER MARCH-LEUBA: I'm glad to hear 15 that.
16 MR. SPRENGEL: To date, we have had 17 several engagements with the NRC staff to talk about 18 our PRA. It is integral to our use of LMP. And one 19 of the benefits, we talked about PRISM before and how 20 PRISM has served as, you know, part of the foundation 21 of the Natrium design. PRISM had a very well-22 developed PRA, and we have used that as a really 23 advanced starting point. So we're not starting from 24 nowhere on our PRA, and we've gone further since that 25 time. I do think our PRA is quite advanced at this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76 1 point, and, of course, we'll go forward with 2 additional steps of peer reviews, and that's kind of 3 going to be going on in parallel to our CPA. But all 4 that to be said that we are moving forward and we are 5 not trying to just submit the very minimum because, at 6 the end of this, we are planning to construct. We are 7 moving right into activities, and so I think our level 8 of detail should be, hopefully, more advanced to meet 9 the needs.
10 Okay. So licensing modernization project, 11 we are using it. Of course, some of the aspects of 12 it, LB identification, SSC classification, and 13 defense-in-depth, that's just part of the fundamentals 14 of what the LMP process outlines. NEI-18-04, serving 15 the basis of that and the staff's endorsement of it.
16 We're also using Reg Guide 1.232, so recently we've 17 submitted our principal design criteria into the staff 18 January of this year. So that is undergoing staff 19 review at this time.
20 And then at the bottom there, building off 21 of what was mentioned earlier, we're following the 22 ARCAP and TCAP activities for the content of our 23 application. And so NEI-21-07 is one of the kind of 24 foundational documents amongst the many, many draft 25 guidance documents that are out there. We are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 1 continuing to follow those and we'll submit under the 2 ARCAP-TCAP kind of structure of the CPA.
3 MEMBER PETTI: So I don't want to put 4 words in your mouth, but it sounds like if Part 53 5 were here today, you've got many of the core elements 6 of Part 53.
7 MR. SPRENGEL: Sure, yes. I think, yes, 8 Part 53 not being here, I guess we haven't gone that 9 far, but yes.
10 MEMBER PETTI: But you've certainly got, 11 I'm going to call it the heart of it.
12 MEMBER MARCH-LEUBA: At ACRS, we fly at a 13 40,000-foot level and trying to see everything. One 14 thing we emphasize a lot is the LB, the licensing 15 basis event selection. It's human nature to start 16 with light water reactor events and scratch out the 17 ones that aren't going to apply to me. What we 18 emphasize a lot is you should not do that. Start with 19 a wide sheet of paper, we hear white sheet of paper a 20 hundred of times while we're here, and think of what 21 could possibly go wrong in your plan that doesn't 22 apply to light waters because it is very human nature 23 to scratch and don't add. So we've been looking into 24 that.
25 MR. SPRENGEL: Thank you for that comment.
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78 1 We haven't started with light water reactor defense, 2 so it is challenging to, you know, to navigate LMP and 3 understand exactly what's a good LB list. But there's 4 still challenges in following the process because it's 5 brand new, but we're using it.
6 MEMBER MARCH-LEUBA: LMP is great. If you 7 forget an event, it could give you the wrong answer.
8 Bad input, bad answer.
9 MEMBER PETTI: The other thing, to 10 piggyback on that, is you've mentioned safety margin 11 because this is sort of first of a kind. You've built 12 out the sodium system, but there's a lot of newness 13 here. And for us, being at 40,000 feet, you know, to 14 have the safety margin buried in some appendix or some 15 technical document supporting the PSAR doesn't do us, 16 you know, as trying to get the public confidence up, 17 where you guys can come up and say a margin here, this 18 is the margin there, you know. That is a really easy.
19 You'll start to see how our letters on advanced 20 reactors look. They look very different than our 21 letters on water reactors purposely because we 22 recognize the first-of-a-kind nature, and so we're 23 looking at things in an LMP Part 53 sort of framework, 24 you know, top-down because these are so new, you know.
25 So I encourage you to make sure that your NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 1 safety margins are clearly articulated when you come 2 to us. It makes our job easier to represent you guys 3 properly.
4 MR. SPRENGEL: Thank you. Looking at our 5 topical reports that we submitted, so there's five 6 listed here. Our quality assurance program 7 description has already been approved. We have a 8 revision that's going through staff review currently 9 in the final stages. I guess, actually, we have the 10 draft SEN in hand right now.
11 George mentioned earlier the energy island 12 interface. That's one that has gone through a staff 13 audit and completed that audit, and so, you know, 14 likely, that would be the next one that would be up 15 for scheduling effort to get in front of the ACRS 16 subcommittee for the Natrium project. I mentioned the 17 principal design criteria here, that was submitted in 18 January of this year. And then the other two are more 19 recently submitted in the last few months, our fuel 20 and control assembly qualification. That's building 21 off of some additional work that we did with the staff 22 and with the Department of Energy on generic advanced 23 fuel qualification, so we used that --
24 MEMBER PETTI: The NUREG that the staff 25 produced, advanced fuel qualification?
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80 1 MR. SPRENGEL: No. In this case, the 2 delivery was actually to the Department of Energy.
3 MEMBER PETTI: Oh, okay. Because you know 4 there's a NUREG-22, I can't remember the number.
5 MR. SPRENGEL: Yes, yes. Those are, yes, 6 they're interlinked in terms of how they were 7 developed absolutely. The topical report we submitted 8 was specific for TerraPower's fuel and control 9 assembly qualification and also touches on things that 10 Eric mentioned that are kind of next generation of 11 fuel that we have on our radar and is important to our 12 overall project.
13 And then our emergency planning zone 14 methodology, which is -- one part we'll see on the 15 upcoming reports, as well. There's other pieces that 16 really fit into that EPZ methodology.
17 MR. BLEY: Before you leave this, Dennis 18 Bley again. I assume you've done something like a 19 PIRT and identified knowledge gaps and things you're 20 working on. If you can say something about that, I'd 21 appreciate it.
22 Also, it appears you've really looked into 23 the salt usage by the solar plants, and you ought to 24 have pretty good reliability data and the like. I 25 don't see any reason why your tanks and piping systems NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 1 would be a whole lot different from theirs, except for 2 maybe temperature. Can you talk about that, too.
3 MR. SPRENGEL: So the first question on 4 PIRTs, we've got, I guess I don't know, I don't want 5 to sound too slang usage here, but we do have lots of 6 PIRTs and in several different topic areas. Recently, 7 I'm trying to think, we met with the staff and 8 presented on our use of PIRTs for our core blockage 9 methodology. We've got some other ones in thermal 10 hydraulic uses. So we've got, you know, PIRTs have 11 been fundamental to how we've looked at what, how we 12 model things and then any gaps that we needed 13 additional testing to support.
14 MEMBER MARCH-LEUBA: Are those likely to 15 be part of the docket, or will it be only internal 16 documents that the staff would have to order you see 17 them? Can I see them eventually?
18 MR. SPRENGEL: I don't want to speak for 19 the staff on that. I'm not sure if PIRTs are part of 20 the basics of what we would submit, so certainly they 21 would be available.
22 MEMBER MARCH-LEUBA: See, the fact is the 23 staff, when they need to see one of your internal 24 documents, they just see it. We at ACRS don't have 25 that. Number one, we compress our review NEAL 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 1 dramatically. And number two, we don't have access to 2 those things. So think about the fact of that. A 3 reading room, one of those famous reading rooms, would 4 help our review.
5 MEMBER PETTI: Other applicants have done 6 that. It's very effective.
7 MEMBER BALLINGER: You don't need a PIRT 8 for a construction permit for the salt side, right?
9 I mean, at what point is the boundary, good old 10 fashioned industrial plant where you can build at risk 11 and you don't need a construction permit for that.
12 MR. SPRENGEL: Yes, I don't think we're 13 doing a PIRT for the energy island. It's just the 14 review of operating experience to understand, you 15 know, taking the technology from a concentrated solar 16 plant that had different requirements on reliability 17 to a nuclear plant that wants a high-capacity factor, 18 is there anything that we need to do or is it good 19 enough? That's the investigation, not really a PIRT.
20 CHAIR REMPE: But we do have to do an 21 environmental assessment for the energy island.
22 MEMBER BALLINGER: For anything. What I 23 mean is you don't need a construction permit from the 24 agency to build the salt side, only to a certain 25 point, right?
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83 1 MR. WILSON: That is correct. This is 2 George Wilson. That is what's in the topical report.
3 It talks about what regulations are applicable or not, 4 so there is a portion of the energy island that would 5 not meet the definition of construction under 50.10, 6 and we would be able to construct it without a limited 7 work authorization.
8 The salt system we're still having the 9 staff look at specifically for portions of it.
10 MR. SPRENGEL: I want to check in.
11 Dennis, I think there was a second part to your 12 question. Did we cover it?
13 MR. BLEY: I think you got it. Oh, yes, 14 I asked have you been able to get pretty good 15 information from the solar folks who use the salt 16 systems, and is there anything really different 17 between your tanks and the things they use?
18 MR. WERNER: Yes, this I Mark Werner.
19 We've been canvassing the concentrated salt power 20 industry quite a bit, and we've got a couple of SNEs 21 on staff that have direct hands-on experience with the 22 systems. So we've been pulling as much knowledge as 23 we can. Being partnered with the DOE through the ARG 24 project has also been helpful because they've got a 25 number efforts through, I think, NREL that we've been NEAL 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 1 able to kind of attach ourselves to.
2 And so we are kind of following some of 3 the tank, I'll call them issues that have been 4 cropping up at different plants and feel like we've 5 got a good plan to address the issues that they're 6 currently seeing. I think one thing that we will 7 bring to the table is, you know, we will probably 8 bring a heightened level of quality to the overall 9 project because it is attached to a nuclear plant. I 10 think that will help out for sure.
11 MR. BLEY: I don't have a clue what 12 temperatures they operate at. Are you a whole lot 13 higher than them? What I'm thinking is is information 14 on system performance and reliability going to be a 15 transferrable from the solar plants over to what 16 you're building?
17 MR. WERNER: Yes, it will be directly 18 applicable. Our temperatures are very close to how 19 they operate. I think they might have a slightly 20 higher hot, and we're kind of down at a slightly 21 colder cool, but we overlap very well.
22 MR. BLEY: That's pretty encouraging. So 23 there's nothing really new here.
24 MR. WILLIAMS: Yes. And the salt that 25 we've chosen for this system is this commercial solar NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 1 salt, so when you hear about solar salt it's a fairly 2 commercial product, and that's what we're using. You 3 know, a lot of the molten salt reactors use a very 4 advanced salt and highly corrosive and whatnot. But 5 this is a well-known salt, and there's lots of it 6 made. It's a commercial product.
7 MR. BLEY: I had a question about the 8 physical nature of the salt. You have really hot 9 liquid salt moving through these pipes. If you get 10 some kind of a break or a leak in a pipe, what's the 11 characteristics of that salt as it comes out? I don't 12 know if salt, if it can flash or what happens to it 13 out there. What's the hazards the operators are going 14 to have to know about and worry about?
15 MR. WILLIAMS: No flashing. It's at 16 atmospheric pressure, so it comes out as a hot liquid 17 and turns into a solid, like that white cake-like 18 substance.
19 MR. BLEY: So pretty benign as it comes 20 out.
21 MR. WILLIAMS: Yes, it is.
22 MR. BLEY: Except if you're touching it.
23 MR. WILLIAMS: Yes, except it's hot. So 24 I should mention, though, our pipes are atmospheric 25 pressure, but they are high temperature. So, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
86 1 know, that definitely involves some design 2 consideration because of that.
3 MEMBER BALLINGER: At 600 C, you're out of 4 the normal stainless steel range. So have you thought 5 about -- the salt people don't use 316 stainless 6 steel. They use other stainless steels.
7 MR. WILLIAMS: Yes, we won't get up that 8 hot.
9 MEMBER BALLINGER: Well, it said 600 C.
10 MR. WILLIAMS: Oh, 600 C, I said stainless 11 steel shows favorable performance with the solar salt 12 up to 600 degrees C. But our core outlet temperature 13 is 510, so nothing is going to be up that hot.
14 MEMBER BALLINGER: I've got 475 stuck in 15 my mind. Certainly, for Section III.
16 MR. WILLIAMS: Yes.
17 MEMBER BALLINGER: Okay, all right. I've 18 got to go back and look and see.
19 MEMBER PETTI: So can you do it in the 20 traditional nuclear part of the code, or do you got to 21 go div 5?
22 MR. WILLIAMS: We go to div 5.
23 MEMBER PETTI: You do go to --
24 MR. WILLIAMS: Yes, we do.
25 MR. SPRENGEL: Looking ahead at our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
87 1 planned topical reports, many of them listed here. We 2 continue to work with the staff on these submittals 3 and getting them to review. This month, we're 4 targeting the HFE program plan document, as well as 5 volcanic hazard assessment. And looking ahead into 6 summer, we've got source term, we talked about 7 relation to our EPZ methodology.
8 Later in the fall, our DBA transient 9 methodology, partial flow blockage, and our code 10 usage. And then in the winter, digital I&C and fuel 11 handling I&C. So lots of activity in terms of getting 12 these into the NRC staff's hands for review this year.
13 MEMBER MARCH-LEUBA: I see you have a 14 topical on stability. Are you worried about 15 oscillations in the core, or is it thermal hydraulic 16 -- what in stability are you worried about?
17 MR. WILLIAMS: It's not VNOM transients; 18 I know that. But probably need to call back to the 19 home office.
20 MR. SPRENGEL: So the stability 21 methodology is, generally, it's describing how the 22 core functions in steady state, so it's kind of a 23 stable reactor.
24 MEMBER MARCH-LEUBA: Not that it's 25 unstable.
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88 1 MR. WILLIAMS: Right.
2 MR. SPRENGEL: Okay. Looking at, we did 3 submit our Natrium engagement plan back in June of 4 2021, coming up on two years of pre-application 5 engagement with the staff, 39 meetings to date. We 6 have seen benefits and have a good working 7 relationship with the NRC staff. And reiterating here 8 our PSAR content is being developed consistent with 9 the ARCAP guidance, so the structure will be different 10 under the ARCAP guidance from NUREG-0800.
11 VICE CHAIR KIRCHNER: How do you view 12 that, Ryan? Is that just a mapping exercise, or are 13 you the first one going through the wicket, or both?
14 MR. SPRENGEL: Yes, both is probably a 15 better way. The guidance in the form of the groups 16 who have done all the guidance between ARCAP and TCAP 17 and it's been spread out into, you know, who kind of 18 has primary, who has the primary lead on developing 19 that guidance, so it's spread out in many documents, 20 and I think all of it is still draft at this point.
21 So it does lay out the structure of what goes where.
22 I guess I'll give an example of where the 23 first time, you know, evolution has come into play.
24 Some of the structure we've modified and we've worked 25 with the staff and, you know, industry stakeholders in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
89 1 terms of some modification of where things might fall 2 just to make sense, primarily from a reviewer and 3 packaging standpoint, and have proposed that that's 4 our path forward, and I think some of that is being 5 incorporated, you know, as the draft finalization.
6 We are also going back and we have 7 reviewed regulations and NUREG-0800, the SRP, for any 8 applicability and any kind of gaps that might be 9 there, and we continue to engage with the staff. You 10 know, when we identify something that we see in 11 expectation from NUREG-0800 but we don't see described 12 kind of one way or the other or maybe lightly in the 13 ARCAP guidance, we reach back out and are looking for 14 clarification on the staff of is that intentional that 15 it was actually removed. Most often, that's not the 16 case. And so then we take that on as an additive of 17 --
18 VICE CHAIR KIRCHNER: That would be my 19 concern because you're doing it as a construction 20 permit, and that's guided by 50.30 and its 21 requirements. ARCAP is coming along later. That's 22 why I said the first order, the mapping exercise. But 23 then whether the staff -- it's something we'll ask the 24 staff, not, you know, are they ready to map NUREG-0800 25 or some version of NUREG-0800 to the ARCAP guidance, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
90 1 so that there's not a lot of -- what I would worry 2 about is not so much that you don't have all the 3 content there, but there's this, like I said, a 4 mapping exercise, and so that the review is efficient 5 for both parties.
6 MR. SPRENGEL: Absolutely. Yes, we are 7 not, we're following the guidance, but we're also 8 doing our due diligence to review the regulations and 9 other sources of kind of, you know, NUREG-0800 is just 10 had more time to develop and be complete and 11 comprehensive, and so we're using the regulation, as 12 well as guidance, to inform and kind of cross-check 13 all those things on our side. I agree it would 14 probably be a fair question for the staff of how 15 they're viewing that, as well.
16 MEMBER BALLINGER: I haven't lost my mind.
17 I'm looking at Table HAA-1130-1 in Division 5, and the 18 upper temperature limit for stainless steel is 425.
19 So I don't know what section we're working to, but 20 Section VIII allows higher temperatures and allows 21 higher allowables. But Division 5,Section III, 22 Division 5, there's a limit there. I saw 600 C on one 23 of your slides, and so maybe I ought to check this.
24 MR. SPRENGEL: Yes, I'll have to check 25 back on the code applicability. I just meant we don't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 1 have anything up at that 600 degree C temperature 2 because the core outlet is at 510. We do have it 3 above 425.
4 MEMBER BALLINGER: I mean, Division 5 is 5 640 pages long, and I'm sure you probably missed 6 something.
7 MS. NEIDER: We'll bring that back to our 8 team and evaluate it.
9 VICE CHAIR KIRCHNER: Members, any 10 questions before we take comments from the public?
11 Okay.
12 MEMBER SUNSERI: Just a question, I guess.
13 I heard a lot of the members talk about Part 53, 14 ARCAP, these kind of things. None of that is approved 15 for us, and they won't be licensing to that, so I 16 don't want to send you all a message that we're 17 holding you accountable to Part 53. Are we? We're 18 not, right? But all we've talked about, we've talked 19 several times about that, so I just want to be clear 20 that's not the standard.
21 VICE CHAIR KIRCHNER: That's where I was 22 going, Matt, that, obviously, they have to use Part 23 50. That's what they're applying for, construction 24 permit under 50.
25 MEMBER BROWN: The point is don't ask NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 1 questions on the other ones.
2 MEMBER SUNSERI: I mean, it was just 3 sounding to me like it could get confusing to the 4 applicant is all.
5 MEMBER MARCH-LEUBA: I'm going to make my 6 typical talk at the end of this presentation based on 7 this comment that we're asking a lot of questions. We 8 don't have any problem with your design. We like your 9 design, and you're doing a great job, and we like this 10 stage of the project. When we want to start 11 scratching on the surface, then we ask questions that 12 you won't like. But right now, thank you for this.
13 Don't take anything we said as meaning anything 14 detrimental.
15 VICE CHAIR KIRCHNER: And these are 16 comments by individual members at this juncture. This 17 is an informational briefing for us, not a critique of 18 the design.
19 Okay. I'd like to open the floor to any 20 comments from the public. If you're participating 21 online, please mute your microphone, state your name, 22 identify your affiliation if relevant, and make your 23 comment, please.
24 MR. MOORE: This is Scott Moore. You may 25 need to press *6.
NEAL 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 1 VICE CHAIR KIRCHNER: Yes, let me see 2 here. We see a hand is up out there. Oh, Ed, yes, go 3 ahead, Ed. Unmute your mike, please, and make your 4 comment.
5 MR. LYMAN: Yes. I thank you. This is Ed 6 Lyman from the Union of Concerned Scientists. I'd 7 just like to point out that we have some serious 8 safety concerns with fast reactors in general and this 9 design in particular. And in the discussion so far in 10 the open session, you only touched on some of the 11 relevant aspects. And what's frustrating is Natrium 12 does not seem to be transparent. When they talk about 13 the difference between sodium-cooled fast reactors and 14 light water reactors, they rarely mention that the 15 time scale for transients is so much shorter in fast 16 reactors that that is a significant safety flaw. And 17 they emphasize the temperature difference between the 18 operating temperature and the boiling point of sodium, 19 but they don't talk about those transients that could 20 lead to a very rapid increase in coolant temperature.
21 And, yes, this reactor does have a positive void 22 coefficient, and, if it does get to a boiling crisis, 23 you could have a very severe reactivity transient.
24 And the absence of a physical containment 25 is a potential significant safety vulnerability they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
94 1 also don't talk about. I know the core assembly 2 events, which I'm sure everyone knows, was a big 3 factor in the Clinch River and FFTF licensing and, as 4 far as I'm concerned, has not been resolved. And so 5 the lack of a strong physical containment is, again, 6 I think a significant open question for this design.
7 So I urge you to explore these questions 8 both in the closed session and also in future open 9 sessions, and that's my comment. Thank you.
10 VICE CHAIR KIRCHNER: Thank you, Ed. Any 11 other members of the public wish to make a comment?
12 Hearing none, then what we will do now is close this 13 Teams' link and go into closed session. Do we need to 14 give any other formalities?
15 CHAIR REMPE: I'd like to suggest that we 16 break until 3:15 to allow the transition to the closed 17 session occur. Also, this is the last that we'll have 18 the open session open today, and so, for those members 19 of the public who want to tune into tomorrow, we'll 20 resume at this link tomorrow at 8:30. Okay.
21 Thank you. And we are recessed.
22 (Whereupon, the above-entitled matter went 23 off the record at 2:54 p.m.)
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
Plant and Licensing Strategy Overview a TerraPower & GE-Hitachi technology NAT-3292 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved.
Purpose
- Provide an overview of the NatriumTM plant design and operation, including the Nuclear Island and Energy Island.
- Describe the innovative features and related research and development activities.
- Discuss the licensing strategy for the Natrium advanced reactor.
Please note, the design is not final. There could be changes to systems, components, plant layout, etc. as the design progresses.
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Control Natrium Safety Features
- Pool-type Metal Fuel SFR with Molten Salt Control Energy Island - Motor-driven control rod runback and
- Metallic fuel and sodium have high compatibility scram follow
- No sodium-water reaction in steam generator
- Gravity-driven control rod scram
- Large thermal inertia enables simplified response to abnormal events - Inherently stable with increased power or temperature
- Simplified Response to Abnormal Events
- Reliable reactor shutdown Cool
- Transition to coolant natural circulation - In-vessel primary sodium heat transport
- Indefinite passive emergency decay heat removal (limited penetrations)
- Low pressure functional containment - Intermediate air cooling natural draft flow
- No reliance on Energy Island for safety functions - Reactor air cooling natural draft flow -
always on
- No Safety-Related Operator Actions or Cool AC power Contain
- Technology Based on U.S. SFR Experience - Low primary and secondary pressure
- EBR-I, EBR-II, FFTF, TREAT - Sodium affinity for radionuclides
- SFR inherent safety characteristics demonstrated - Multiple radionuclides retention boundaries through testing in EBR-II and FFTF Contain SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 3
Key Differences from Light-Water Reactors Leverage inherent features:
- Compact systems, less nuclear sprawl
- Low pressure
- Efficient heat transfer
- Pool design with large coolant inventory
- Modularity
- Parallel construction
- Emergency Planning Zone reduced SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 4
Single Unit Site Steam Generation Turbine Building Energy Storage Tanks Fuel Handling Building Rx Building Standby Diesels Switchyard Salt Piping Rx Aux. Building Control Building Warehouse Firewater
& Admin SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved.
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Reactor Building Reactor Aux. Building Fuel Handling Building Reactor Air Cooling Ducts Refueling Access Area Intermediate Salt Piping to/from Air Cooling Thermal Storage System Sodium Int. loop Ground Level Sodium/Salt HXs Spent Fuel Pool (water)
Intermediate Sodium Hot Leg Head Access Area Intermediate Sodium Cold Leg Reactor and Core Reactor Air Cooling / Reactor Cavity SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 6
Energy Island Cold Molten Salt Thermal Storage Tank Thermal Storage Hot Molten Salt Thermal Storage Tank Thermal Storage Molten Salt Pipe Racks
- Number of tanks based on customers energy need Live steam
- Steam generator trains Steam drum Hot salt from Cold salt to based on size of turbines reactor reactor Superheater Cold salt
- Turbine size based on Hot salt pump Preheater pump customers power need Evaporator Attemperation Feedwater pump Reheater Cold reheat Hot Tank Cold Tank Hot reheat SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 7
Benefits of Sodium Coolant
- High heat capacity - moderate flow rate and easy decay heat removal
- High heat transfer - small core and easy decay heat removal
- High boiling point 883°C (1,621°F) - atmospheric pressure
- Low melting point 98°C (208°F) - practical
- Lack of corrosion LWR oxide fuel BP = Boiling Point (atmospheric)
- Limited auxiliaries PCT 1100 C Natrium metallic fuel BP 345 C @ 155 atm BP 883 C @1 atm CT 300 C PCT 650 C CT 550 C
- Sodium inventory - 800 m3 in reactor SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 8
Sodium in Liquid and Solid States SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 9
Molten Salt for Energy Storage
- 60 NaNO3- 40 KNO3
- Molten salt inventory - 30,000 tons
- Gross tank energy storage capacity with Type 1 fuel - 1,971 MWh Benefits of using molten salt:
- Long design life with negligible performance degradation
- Temperature range 238 - 621°C (460 - 1,149 °F)
- High thermal energy storage efficiency (~99%)
- Readily available due to its common use for heat storage and in solar plants
- Relatively low levelized cost of energy at grid scale compared to battery storage SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 10
Salt in Solid and Molten States SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 11
Example of Simplified Nuclear Systems LWR Emergency Core Cooling Natrium Reactor Air Cooling
- High Pressure Injection (1000+ PSI)
- Atmospheric Pressure (<1 PSI)
- Large Water Inventory Requirements
- Unlimited Air-Cooled Heat Sink Supply
- Active Valve and Pump Operation
- Fully Passive (Always in Operation)
- Multiple Trains and Sub-systems
- Singular Rugged System 9799218-13_r0 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 12
Key Features of Reactor Equipment Design CRDMs Heat Transport In-Vessel Transfer Loop Piping Machine
- Pool Type Integral Reactor
- Large Volume of Sodium Hot Pool Guard Vessel Coolant
- Atmospheric Pressure Heat Exchanger Pump
- Separation of Hot and Cold Reactor Vessel Pool Regions
- Mechanical Pumps Reactor Core
- In-vessel Refueling Cold Pool Core Inlet Plenum SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 13
Heat Removal Normal
- Intermediate Air Cooling Normal Shutdown Heat Removal Forced Flow Passive
- Intermediate Air-Cooling Non-Safety-Related Heat Removal Natural Draft Flow Inherent
- Reactor Air Cooling Safety-Related Heat Removal Natural Draft Flow - Always On SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 14
Overview of the 10 CFR Part 50 Licensing Process for the Natrium Advanced Reactor Submit NRC Safety Review Mandatory Construction Operating Application for ACRS Fuel Load 10 CFR Parts 20, ACRS Review Hearings Permit License Operating License Review and Start Up NRC holds 50, 73, and 100 (Comm./ASLB) Issued Issued (FSAR)
Public Meeting at the Site Environmental Review 10 CFR Part 51 Submit NRC Application for Construction Permit
( ex PSAR and ER)
Key:
Opportunity for Public Participation Milestone Activity SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 15
Proposed Application of LMP
- Use of LMP for the Natrium design:
- Regulatory Guide 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors
- NEI 18-04, Risk-Informed Performance-Based Technology Guidance for Non-Light Water Reactors
- LMP analysis, including LBE selection, plant-level SSC classification (input to system-level SSC classification) and evaluation of defense-in-depth adequacy.
- NEI 21-07, Technology Inclusive Guidance for Non-Light Water Reactors describes SAR content for advanced reactors using NEI 18-04 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 16
Topical Reports Submitted
- Quality Assurance Program Description
- Natrium Nuclear Island and Energy Island Interface
- Principal Design Criteria for the Natrium Advanced Reactor
- TerraPowers Fuel and Control Assembly Qualification
- Emergency Planning Zone Methodology SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 17
Topical Reports Planned
- HFE Program Plan and Methodologies
- Volcanic Hazards Assessment
- Mechanistic Source Term Methodology
- Radiological Release Consequences Methodology
- Partial Flow Blockage Methodology
- Engineering Computer Codes for the Natrium Reactor
- Reactor Stability Methodology
- Defense-in-Depth and Diversity I&C Strategy
- Digital I&C (Architecture and Design)
- Fuel Handling Instrumentation and Control SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 18
Path Forward
- Regulatory Engagement Plan - June 2021
- 39 Pre-application meetings held to date
- PSAR content is being developed consistent with draft Advanced Reactor Content of Application Project (ARCAP) guidance SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 19
Questions?
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Acronym List ACRS - Advisory Committee on Reactor Safeguards IAC - intermediate air cooling system ARCAP - Advanced Reactor Content of Application Project LBE - licensing basis event ASLB - Atomic Safety and Licensing Board LMP - Licensing Modernization Project ASME - American Society of Mechanical Engineers LWR - light-water reactor BP - boiling point MWh - megawatt-hour CFR - Code of Federal Regulations NEI - Nuclear Energy Institute CRDM - control rod drive mechanism NRC - U.S. Nuclear Regulatory Commission CT - cladding temperature PCT - peak cladding temperature DBA - design basis accident PSAR - Preliminary Safety Analysis Report EBR - Experimental Breeder Reactor PSI - pounds per square inch ER - Environmental Report RAC - reactor air cooling system FFTF - Fast Flux Test Facility SAR - Safety Analysis Report FSAR - Final Safety Analysis Report SFR - sodium-cooled fast reactor HFE - human factors engineering SSC - structure, system, and component HX - heat exchanger TREAT - Transient Reactor Test Facility I&C - instrumentation and control SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 21
ENCLOSURE 3 Plant Overview Presentation Material - Closed Meeting Non-Proprietary (Public)
Plant Overview a TerraPower & GE-Hitachi technology NAT-3293 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved.
Portions of this presentation are considered export controlled information (ECI). ECI can be disclosed to Foreign Nationals only in accordance with the requirements of 15 CFR 730 and 10 CFR 810, as applicable.
Portions of this presentation are considered proprietary and TerraPower, LLC requests it be withheld from public disclosure under the provisions of 10 CFR 2.390(a)(4).
Nonproprietary versions of this presentation indicate the redaction of such information using (( ))(a)(4), (( ))ECI, or (( ))(a)(4), ECI.
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Topics
- Reactor Core Design
- Reactor Equipment
- Refueling Equipment
- Heat Removal Systems
- Sodium Leak Protection/Mitigation
- Functional Containment Strategy
- Source Term and Emergency Planning Zone Strategy
- Energy Island Systems
- Reactor Protection System Parameters
- Representative Transients:
- Basic SCRAM Design Transient
- Basic Runback Design Transient
- Uncontrolled Rod Withdrawal Design Basis Accident (DBA)
- Loss of Offsite Power DBA SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 3
Reactor Core Overview Core Assembly Reactor Core Map Handling Load pads Socket Duct Inlet Nozzle Reactor Core SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 4
Reactivity Feedback Mechanisms of SFRs
- Doppler feedback: Effect of changes in neutron fission and absorption cross sections due to Doppler broadening
- Negative at temperatures above normal
- Core radial expansion: Due to thermal expansion and irradiation-induced swelling
- Negative at temperatures above normal due to enhanced leakage and core locked
- Fuel axial expansion: Effect of thermal expansion and transient swelling of especially the metallic fuels (and cladding)
- Negative at temperatures above normal due to reduced number density of fissionable isotopes
- Coolant density and void worth: Effect of changes in coolant density at elevated temperatures
- Can be positive due to reduced sodium moderation/absorption, or negative due to enhanced neutron leakage
- Control rod driveline expansion: Due to difference in thermal expansion of control rod driveline and reactor vessel
- Can be positive or negative depending on expansion relative to reactor vessel expansion SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 5
Fuel Types
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((
Pin Strip Layer Assembly
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Key Features of Reactor Equipment Design CRDMs Heat Transport In-Vessel Transfer Loop Piping Machine
- Pool Type Integral Reactor
- Large Volume of Sodium Hot Pool Guard Vessel Coolant
- Atmospheric Pressure Heat Exchanger Pump
- Separation of Hot and Cold Reactor Vessel Pool Regions
- Mechanical Pumps Reactor Core
- In-vessel Refueling Cold Pool Core Inlet Plenum SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 8
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In-Vessel Refueling
((
Refueling Eq uipment K ey Features:
- Remote, in-vessel equipment
- Drives located on head
- Access to all in-vessel locations
- All core components have same interface
- Multiple degrees of freedom
- New Fuel
- Into reactor vessel (RV) using lift
- Into core using in-vessel transfer machine
- Spent Fuel
- From Core to in-vessel storage
- From storage to lift
- Leave RV using lift
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In-Vessel Transfer Machine Operations
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Ex-Vessel Fuel Handling Process Overview
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Ex-Vessel Fuel Handling Process Equipment
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Intermediate Heat Transport Loop
((
Key Equipment:
- Sodium/Salt Heat Exchanger (HX)
- Nitrate Salt Sodium
- Intermediate Pumps
- pumps similar to primary sodium pump (PSP)
- Intermediate Air Cooling
- Provide local heat rejection during start up and shutdown
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Heat Transport Architecture NI Control hot
- Cold pump speed
- Cold tank level Power Cycle Loop tank
- Cold salt temperature (Water) cold tank EI Salt System (Nitrate Salt)
Commercial Construction Primary Pool Graded Approach:
(Sodium) Constructed in EI Control Reactor accordance with
- Hot pump speed NQA
- Hot tank level 9799218-7b_r0
- Hot salt temperature SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 18
((
Reactor Air Cooling
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Heat Removal Normal
- Intermediate Air Cooling Normal Shutdown Heat Removal Forced Flow Passive
- Intermediate Air-Cooling Non-Safety-Related Heat Removal Natural Draft Flow Inherent
- Reactor Air Cooling Safety-Related Heat Removal Natural Draft Flow - Always On SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 20
Sodium Leak and Fire Protection
((
Prevention & Lessons Learned
- Significantly reduced quantity of sodium piping
- Leak Jacketing / Guard piping
- Remove sodium to steam interface
- Mitigation features:
- Reactor vessel surrounded by guard vessel
- Inerted
- Leak and fire detection
- Reactor Head / Head Access Area
- Steel lined cells or leak jacketing
- Inerted
- Leak and fire detection
- Leak protection to not compromise SR functions
- Cells containing Intermediate sodium will have
- Catch pans with suppression plates at critical areas
- Leak and fire detection )) (a)(4)
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Functional Containment - Diverse Barriers The Natrium' design is well suited for Functional Containment:
- Low operating pressures & large margin to sodium boiling
- Low differential pressure between vessel and compartments
- High conductivity coolant & passive emergency core cooling
- Design precludes any consequential Loss of Coolant Accident (LOCA)
Environment Diverse Barriers in Natrium Functional Containment:
Fuel Environment Reactor Handling Head Building* Building*
Fuel Sodium Cover Matrix Coolant Gas Access Area Reactor Auxiliary Building* Environment
- Sub-compartments within building provide additional confinement, depending on event.
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Functional Containment - Analysis and Source Term
- Quantify compartment to compartment leakage (P/T dependent).
- Assess aerosol behavior in compartments (deposition/condensation, radionuclide decay, and agglomeration)*.
- Assess sodium-chemical reactions in air-filled spaces (event specific).
- Assess barrier performance for licensing basis event with radiological consequences, and Design Basis Accidents (includes cliff edge effects, considerations for severe accidents, EPZ methodology, and others).
- phenomena also considered in the cover gas region Adapted from ANL-ART-49 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 23
Site Boundary SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE- DE-NE0009054 NE0009054 Copyright© 2023 TerraPower TerraPower, LLC. All Rights Reserved. 24 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All rights reserved.
Load Following w/ Integrated Energy (Thermal) Storage
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Load Following w/ Integrated Energy (Thermal) Storage
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((
Energy Island Capacity Optimization
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Energy Island -
Thermal Storage System
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Energy Island - Steam Generator Equipment
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Energy Island -
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Turbine, Generator and Feedwater Systems
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Natrium Hybrid Main Control Room -
Nuclear Island + Energy Island Human Machine Interfaces
((
- Nuclear island control system (NIC) operations independent from energy island control system (EIC) operations
- Group view display system (GVDS)
- No safety-related (SR) action initiated from the main control room (MCR)
- Power runback (DL2) can be manually initiated or stopped by operators for operation flexibility
- Manual reactor trip or primary sodium pump trip can be initiated by an operator (for DID purpose only)
- Fuel handling control room located in the fuel handing building is independent of MCR
- All plant parameters (SR, or non-safety) available on NIC video display units for ease of operation.
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RPS Reactor Trip Parameters Note: Current trip setpoints are under continued development and do not reflect the final selection of parameters and inputs for the plant design or licensing basis.
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Basic SCRAM Design Transient Sequence of events following a SCRAM function:
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Basic SCRAM Design Transient (continued)
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Basic SCRAM Design Transient (continued)
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Basic Runback Design Transient Sequence of events following a RUNBACK function:
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Basic Runback Design Transient (continued)
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Uncontrolled Rod Withdrawal
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Uncontrolled Rod Withdrawal
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((
Uncontrolled Rod Withdrawal
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((
Uncontrolled Rod Withdrawal
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Uncontrolled Rod Withdrawal
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Loss of Offsite Power
- Event sequence (long term response):
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Acronym List AHX - sodium-air heat exchanger IVTM - in-vessel transfer machine ANL - Argonne National Laboratory LOCA - loss of coolant accident BLTC - bottom loaded transfer cask MCR - main control room CCCS - core component conditioning station NI - nuclear island CFR - Code of Federal Regulations NIC - nuclear island control system CRDL - control rod drive line NQA - Nuclear Quality Assurance CRDM - control rod drive mechanism NSS - nuclear island salt heat transport system DBA - design basis accident P/F - power to flow DID - defense-in-depth PIE - post-irradiation examination DL - defense line PRC - pin removal cell ECI - export controlled information PSP - primary sodium pump EI - energy island RAC - reactor air cooling system EIC - energy island control system RPS - reactor protection system EPZ - emergency planning zone RBFP - reactor building floor plug ESS - energy island salt heat transport system RIS - reactor instrumentation system EVHM - ex-vessel handling machine RV - reactor vessel EVST - ex-vessel storage tank SFR - sodium-cooled fast reactor FTP - fuel transfer port SR - safety-related GVDS - group view display system XIS - nuclear instrumentation system HX - heat exchanger IAC - intermediate air cooling system IHT - intermediate heat transport system ISP - intermediate sodium pump SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright© 2023 TerraPower, LLC. All Rights Reserved. 51