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
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 704TH MEETING 4
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5
(ACRS) 6
+ + + + +
7 OPEN SESSION 8
+ + + + +
9 THURSDAY 10 APRIL 6, 2023 11
+ + + + +
12 The Advisory Committee met via hybrid In-13 Person and Video-Teleconference, at 8:30 a.m. EDT, Joy 14 L. Rempe, Chairman, presiding.
15 16 COMMITTEE MEMBERS:
17 JOY L. REMPE, Chairman 18 WALTER L. KIRCHNER, Vice Chairman 19 DAVID A. PETTI, Member-at-Large 20 RONALD G. BALLINGER, Member 21 CHARLES H. BROWN, JR., Member 22 VICKI M. BIER, Member 23 VESNA B. DIMITRIJEVIC, Member 24 GREGORY H. HALNON, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 JOSE MARCH-LEUBA, Member 1
MATTHEW W. SUNSERI, Member 2
3 ACRS CONSULTANT:
4 DENNIS BLEY 5
7 DESIGNATED FEDERAL OFFICIAL:
LARRY BURKHART 10 11 ALSO PRESENT:
12 JOE ASHCRAFT, NRR 13 ERIC BENNER, NRR 14 GILBERTO BLAS RODRIGUEZ, NRR 15 SAMIR DARBALI, NRR 16 WILLIAM JESSUP, NRR 17 CHRIS LEVESQUE, TerraPower 18 ED LYMAN, Public Participant 19 KHOI NGUYEN, NRR 20 TARA NEIDER, TerraPower 21 JASON PAIGE, NRR 22 RYAN SPRENGEL, TerraPower 23 RICHARD STATTEL, NRR 24 DINESH TANEJA, NRR 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 MARK WERNER, TerraPower 1
ERIC WILLIAMS, TerraPower 2
GEORGE WILSON, TerraPower 3
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 P R O C E E D I N G S 1
8:30 a.m.
2 CHAIR REMPE: Good morning. It is 8:30 on 3
the East Coast, and this meeting will now come to 4
order. This is the second day of the 704th Meeting of 5
the Advisory Committee on Reactor Safeguards.
6 I'm Joy Rempe, Chairman of the ACRS.
7 Other members in attendance are Ron Ballinger, Vicki 8
Bier, Charles Brown, Vesna Dimitrijevic, Greg Halnon, 9
Walt Kirchner, Jose March-Leuba, Dave Petti, and Matt 10 Sunseri. We do have a quorum.
11 Similar to yesterday, the Committee is 12 meeting in person and virtually. A communications 13 channel has been opened to allow members of the public 14 to monitor the Committee discussion. Mr. Larry 15 Burkhart is the Designated Federal Officer for today's 16 meeting.
17 During today's meeting, the Committee will 18 consider the following topics: planning and procedures 19 session and Commission meeting preparation, TerraPower 20 and Natrium reactor design overview and 3D model 21 walkthrough. I note that portions of our discussions 22 on both of these items may be closed.
23 The transcript of the open portions of the 24 discussion on topic two is being kept, and it's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 requested that speakers identify themselves and speak 1
with sufficient clarity and volume so they can be 2
readily heard. Additionally, participants should mute 3
themselves when they're not speaking.
4 At this time, I'd like to ask other 5
members if they have any opening remarks. Not hearing 6
any, then I'd like to go off the record at this time 7
before we start with topic one today. And I'd also 8
ask the court reporter to return at 1 p.m. for our 9
discussion on topic two.
10 (Whereupon, the above-entitled matter went 11 off the record at 8:31 a.m. and then went back on the 12 record at 1:00 p.m.)
13 CHAIR REMPE: Okay. It's 1 p.m. on the 14 East Coast, and we are going to reconvene, and I'm 15 going to ask Member Kirchner to lead us through the 16 second topic for today. Walt.
17 VICE CHAIR KIRCHNER: Okay. We have 18 guests here today with us from TerraPower, and they 19 are going to give us an informational presentation on 20 the Natrium design they are developing and give us an 21 overview. I'd ask Chris Levesque to perhaps somewhere 22 through this address the status of their engagement 23 with the staff, which is useful information for us 24 when we look at our longer-range planning going 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
6 forward.
1 So with that, Chris, I'm just going to 2
turn it over to you to introduce your team. And thank 3
you for being here. We like the engagement in person, 4
so, hopefully, you'll like the engagement with us in 5
person, as well. Go ahead.
6 MR. LEVESQUE: Thank you, Madam Chair, Mr.
7 Vice Chair, and thanks to the entire Advisory 8
Committee on Reactor Safeguards at this time. We look 9
forward to the engagement today and to multiple 10 engagements with you in the future as we go forward 11 with the Natrium licensing process.
12 I'm joined by Tara Neider, who is the 13 project director for the Natrium project, Senior Vice 14 President and project director at TerraPower. She 15 oversees the entire Natrium project, including the 800 16 design engineers who are working on the project today 17 in the procurement and instruction processes, as well.
18 I think I might defer further intros from 19 the TerraPower team because we'll have multiple people 20 coming up today, and we'll save those intros for when 21 those subject matter experts come in.
22 I wanted to begin by sharing, you know, 23 the common value, the common understanding that we at 24 TerraPower feel towards reactor safety. We believe 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 the Natrium design, you know, addresses the NRC's 1
requirements and the engineering requirements that are 2
needed for reactor safety. It is a Gen IV reactor, 3
which we expect Gen IV reactors to have even higher 4
margins for safety and we'll utilize excessively safe 5
and inherent systems to achieve safety.
6 MR. NGUYEN: Excuse me. Could you 7
identify yourself for the court reporter? Thank you.
8 MR. LEVESQUE: Yes. My name is Chris 9
Levesque, President and CEO of TerraPower.
10 We're also looking forward, in addition to 11 the engagement with the ACRS, we're closely monitoring 12 the development of the Advanced Reactor Content of 13 Application Project. We know, following the Nuclear 14 Energy Innovation and Modernization Act, there's 15 various developments going on in regulation, so we're 16 watching these closely. But we are making this first 17 application under 10 CFR 50, an established statute.
18 We're also following the NRC's guidance on 19 the pre-application engagement, and we have 20 considerable pre-application engagement already behind 21 us with 39 meetings in pre-application, 5 topical 22 reports, and 3 white papers. And that engagement with 23 the NRC has been very constructive. We look forward 24 to engagements like this with the ACRS and engagements 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 with the NRC in the pre-application review as a way of 1
improving our eventual construction permit application 2
and operating license application.
3 Knowing that Natrium is a Gen IV design 4
and most of the experience in the U.S. is in the light 5
water arena, we're doing as much as we can to 6
familiarize the staff and other stakeholders on sodium 7
fast reactors. Best example of that is the multiple 8
training sessions we've given for NRC staff on sodium 9
fast reactors and the attributes of Natrium.
10 I want to finish by noting that this first 11 Natrium plant, which will be built in the state of 12 Wyoming at a retiring coal plant, is being built in 13 conjunction with a very important public-private 14 partnership, the Advanced Reactor Demonstration 15 Program. Both TerraPower and X-energy were recipients 16 of the ARDP Demonstration Project Award. That award 17 involves the Department of Energy funding half of the 18 costs of these first projects, and that's intended to 19 help us overcome many first-time costs associated with 20 the first license, the first design, learning curve, 21 frankly, because the U.S. has somewhat fallen out of 22 experience with new builds.
23 So the ARDP, we feel, is a very, very 24 important national program. We're proud to be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 recipients of that program, and a very important part 1
of that program is that the first owner of the Natrium 2
nuclear power plant will be a commercial owner.
3 So we believe that public-private 4
partnerships are very important. We're competing on 5
the world stage with Chinese and Russian-state owned 6
entities. We believe that American innovation and 7
American public-private partnerships are really the 8
way to compete with the state-owned entities, and we 9
look forward to engagement with all of our government 10 stakeholders.
11 I'm going to now hand it off to Tara 12 Neider, again, our Senior Vice President and project 13 director for Natrium.
14 CHAIR REMPE: So I don't know if this is 15 your first time at ACRS, but we like to ask questions.
16 MR. LEVESQUE: Please.
17 CHAIR REMPE: Are we allowed to ask you 18 questions? I know you've got a meeting you've got to 19 go --
20 MR. LEVESQUE: Absolutely. No, we look 21 forward to that. Yes.
22 CHAIR REMPE: I'm just curious. I mean, 23 you've mentioned that you've got five topical reports 24 in the queue and you mentioned that you look forward 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
10 to future engagements with ACRS. Do you have a time 1
frame when you think you'll be back before ACRS?
2 MR. LEVESQUE: Well, we'll share, in 3
further reports, we'll share our plans for our 4
construction permit application, which will be early 5
next year. We would plan multiple meetings. We 6
certainly don't think it's too late or too early to be 7
engaging you because we have a fully mobilized 8
project. I know there's many projects that you can 9
read about everyday in the U.S., but I would hazard to 10 say maybe none of them have 800 design engineers 11 working on a design, soil borings at the site already 12 complete. I mean, this project is quite advanced and, 13 given the construction permit application early next 14 year, we think this is a great time for the first ACRS 15 meeting.
16 George, Ryan, any --
17 MR. WILSON: This is George Wilson. I'm 18 the VP of Regulatory Affairs for TerraPower. Right 19 now, Billy Jessup is in the room, we have scheduled 20 meetings --
21 MR. BLEY: Can you talk more into the 22 microphone?
23 CHAIR REMPE: You need to be closer to the 24 mike for the people online. I'm sorry.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 MR. WILSON: This is George Wilson, 1
TerraPower, VP of Regulatory Affairs. I know that we 2
have some topical report meetings lined up with the 3
ACRS starting in August, this August. But we wanted 4
to come in, give an overview now, so that you guys 5
could potentially ask questions and then we can, you 6
know, as we go from there.
7 CHAIR REMPE: I'd caution you that it's 8
not showing up, we have our schedule that we went 9
through earlier today, and it's not on the schedule.
10 So please work with our staff to make sure that it is 11 as time that members know in advance, and they can 12 make their plans accordingly if you're going to be 13 coming in in August. That's why I asked --
14 MR. WILSON: That was the feedback that we 15 had gotten from NRR. I think Branch Chief Billy 16 Jessup is here, if he wants to add some more.
17 CHAIR REMPE: You can just stand and talk.
18 We've gone with new technology. And be sure and say 19 your name.
20 MR. JESSUP: Good afternoon. This is Bill 21 Jessup, Chief of Advanced Reactor Licensing Branch 1, 22 NRR. Our project managers are working very closely 23 with ACRS staff as we progress through the review of 24 the various topical reports and white papers, as well.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
12 I know the ACRS is not looking at white papers. We 1
are working very closely with ACRS staff to schedule 2
forthcoming meetings that have been --
3 CHAIR REMPE: It's not on the August 4
schedule. If we're going to meet in August, be sure 5
and get it on the schedule, but that helps to know.
6 And at some point, I know some of this proprietary, 7
but some of your, if you can tell us which report it 8
is that's going to be coming in in August, that would 9
be helpful for us, too, to know. I don't know if you 10 want to say it on the record here.
11 MR. WILSON: I think the NRC staff, we 12 have a -- this is George Wilson again. We have a 13 topical report. Our reactor design is a little bit 14 different. We have a topical report on the nuclear 15 island, energy island interfaces. Since we can 16 actually operate the turbine, it's not dependent, we 17 can operate the turbine at, like, 100-percent power 18 and the reactor at 10-percent power. They're not 19 linked like a light water reactor, so, based on that, 20 there are certain regulations that will not apply to 21 the Natrium design. So that is a topical report right 22 now the NRC staff has completed the audit on and 23 they're finishing up their review, and I know that 24 will be one that's coming to ACRS.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 We submitted principal design criteria.
1 That will be another one that the NRC has just started 2
reviewing. So those are ones that will be coming up 3
to the ACRS first.
4 CHAIR REMPE: Okay. I did look over your 5
regulatory engagement plan, but I just couldn't get a 6
clue of what's coming in and when. So thank you.
7 MS. NEIDER: As was mentioned, I'm Tara 8
Neider. I'm the project --
9 MEMBER PETTI: Microphone, please.
10 MS. NEIDER: Oh, okay. As Chris 11 mentioned, I'm Tara Neider. I'm the project director 12 for Natrium, and I wanted to talk a little bit about 13 the Kemmerer site that we're selecting to build this 14 plant on and then discuss a little bit about the team 15 that we've put together, which I think is a pretty 16 world class team.
17 As Chris mentioned, we are building the 18 Natrium reactor at the site of a coal plant that is 19 scheduled to shut down. And I can tell you that when 20 we did our tour of various places to put the plant in 21 Wyoming, this town, actually there's two towns, 22 Diamondville and Kemmerer, they really welcomed us 23 with open arms. It's really important for this 24 community to have something to replace the work that 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 was done at both the coal mine, you know, which feeds 1
the coal plant, and then the coal plant. So it's been 2
a very, very cooperative agreement we've had with the 3
town. However, it is a small town, so we do have to 4
make sure that we can be prepared for the number of 5
people that we have which will be coming during the 6
construction to build the plant.
7 The plant is a sodium fast reactor, and 8
it's a pool type reactor, and most of the safety-9 related systems are underground. Mark and Eric will 10 be talking later about the details of that design.
11 But what really puts us apart from others is that the 12 power generated from Natrium is dispatchable. Between 13 the nuclear island and the energy island, nuclear 14 island can run all the time, the energy island can 15 actually go up and down in power as the power is 16 needed on the grid.
17 Our team is, as I said, a world class 18 team. Our design partner is GE-Hitachi. They have 19 very strong experience in sodium fast reactors and, 20 obviously, commercial BWR plants. And we also have 21 Bechtel. Bechtel is doing the construction, and 22 they're also responsible for the design of the energy 23 island and some of the civil works on the nuclear 24 island.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 We also have a lot of other entities that 1
are part of our team. Specifically, we've engaged 2
with a number of nuclear utilities to provide us 3
guidance on the operations piece and make sure that 4
both operations and maintenance is considered in our 5
design. Those include Duke Energy and Energy 6
Northwest, as well as Pacific Corp, who will be the 7
owner/operator of the plant eventually. Pacific Corp 8
or, actually, a subsidiary of Rocky Mountain Power is 9
the utility in that area, and they will eventually own 10 and operate the plant, so they are part of our team 11 moving forward.
12 We also have a number of national labs 13 that are doing various support with regard to modes 14 and methods and also testing of our advanced fuel for 15 this reactor. And we have a number of other entities, 16 as well, a number of universities. The University of 17 Wisconsin, Oregon, and also NC State is part of our 18 team. So we really feel that we've put together a 19 pretty top-notch team to be able to move forward.
20 In addition to our teammates, we are 21 trying to make sure that our plant meets the needs of 22 the commercial market. And as a result of that, we've 23 formed what we call the Natrium utility advisory 24 group, but we also now call it users because there's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 a number of companies in the advisory committee that 1
are not utilities at all. They're industry 2
participants. And so those people we meet with on a 3
regular basis, and they provide us guidance and take 4
a look at how we're going through things. And I think 5
we're up to about 30 people, 30 companies on that 6
advisory group, so it's very positive.
7 We just joined INPO as just a supplier at 8
this point, but we think that having joined INPO will 9
actually provide us a lot of experience, especially 10 the lessons learned from the industry going forward.
11 And then, finally, we also have a memorandum of 12 understanding with JAEA, the Japanese Atomic Energy 13 Agency, and Mitsubishi. The reason that we have that 14 memorandum in place is that they will be providing 15 their own sodium fast reactors, but they also have a 16 fair amount of experience running Joyo and Monju in 17 Japan, so we really felt that their guidance would 18 help us with this plant. And where they've helped us 19 if they've participated in some of our design reviews 20 going forward.
21 So those are my main points I wanted to 22 make today, and I think, at this point, we'll ask if 23 you have any questions or comments before we get into 24 the technical side.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 MR. BLEY: This is Dennis Bley. You've 1
said you joined INPO. Are you considering being 2
involved with FLEX and SAFER, or does that just not 3
match up with you at all?
4 MS. NEIDER: We are considering FLEX and 5
SAFER, yes. Not SAFER. FLEX, I know, we are 6
considering a FLEX building, but it's going to depend 7
on where our analysis comes up with things.
8 MR. BLEY: Thank you.
9 VICE CHAIR KIRCHNER: Maybe we'll get into 10 this as we go forward, but, just at a very high level, 11 just the first order, is this an evolution of a PRISM 12 design that GE had been advocating back in the mid -
13 late 80s time frame?
14 MS. NEIDER: There are certain elements of 15 the PRISM design. We kind of combined the TerraPower 16 traveling wave reactor and the GE PRISM reactor to 17 develop the Natrium reactor. The reason, you know, 18 the main change from those two is the size, but we did 19 find some things on the various components that we did 20 have to do differently just because --
21 VICE CHAIR KIRCHNER: When we get into 22 closed session, maybe just keep this in mind. I'd 23 like to ask some questions about what you learned from 24 the PRISM design. In particular, the NRC back in that 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
18 time frame, actually, a CER on the PRISM reactor, so 1
it did get a fair amount of attention --
2 MS. NEIDER: Yes. And that safety 3
evaluation report has been combed over by our 4
engineers, and there were certain open items, like 5
transient testing, that was required, and we've added 6
that to our program. So we do believe those things 7
are covered.
8 VICE CHAIR KIRCHNER: Good. Is this going 9
to be the end of your open presentation, or are you 10 going to shift gears here?
11 MR. WILLIAMS: We haven't actually started 12 our open presentation.
13 (Laughter.)
14 VICE CHAIR KIRCHNER: Well, it's not 15 unusual at ACRS to not get beyond the first --
16 (Laughter.)
17 VICE CHAIR KIRCHNER: -- so I'm just 18 checking. Okay. Go ahead.
19 MR. WILLIAMS: Okay. I'll introduce 20 myself real quick. My name is Eric Williams. I'm a 21 Senior Vice President at TerraPower, and I'm the 22 design authority for the Natrium project. And I'm 23 going to be talking today about a plant design 24 overview with three main points really to provide an 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 overview of the plant design and operation, including 1
both the nuclear island and the energy island. We're 2
going to be focusing on more of the innovative 3
features that we have, you know, comparing us to a 4
light water reactor or something operating today, what 5
would be the key differences for us, and also some 6
differences between historical sodium fast reactors 7
and the Natrium reactor, as well. And then I'll hand 8
it off to Ryan, and he'll talk about the licensing 9
strategy.
10 So this first slide to begin with, I'm 11 going to spend a little bit of time on this one.
12 We've talked about this one in all of our NRC 13 engagements because it really has the key safety 14 features all contained on one slide.
15 And so when you look at the Natrium 16 reactor, it is a 840 megawatt thermal core output, and 17 the key innovative features are really in that very 18 first bullet. It's a pool type reactor, as opposed to 19 a loop style reactor, so you have a large reactor 20 vessel with a large liquid inventory of sodium within 21 it. There are free surfaces within this reactor 22 vessel, so it's a little different way of thinking 23 from a loop reactor. It also uses metal fuel, a type 24 of fuel that was undergoing heavy amounts of research 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 and development in the U.S. legacy SFR program but 1
never got turned into a full metal core until now. So 2
there's various reactors out there that have used 3
oxide fuel. I think even PRISM was intended to, you 4
know, start out with some oxide fuel and convert to 5
metal fuel. Natrium is going to start out with all 6
metal fuel based on the research and development done 7
at EBR-II and FFTF on this metal fuel type.
8 And then it's got a molten salt energy 9
island, so that's definitely a new piece that enables 10 us to store thermal energy in the energy island coming 11 off the nuclear island and use that to generate 12 electricity at a rate demanded by the grid, as Tara 13 mentioned in her introduction. And it also provides 14 another source of large liquid inventory that also 15 adds, along with the liquid in the reactor vessel, 16 sort of a buffer to any, you know, immediate 17 challenges to core heat removal.
18 So due to those things, we'll get into 19 that a little bit more as we go through these slides, 20 and you'll see the effects of those major features.
21 But those are the three big ones.
22 And a couple of other things about the 23 metal fuel to keep in mind, it has a high degree of 24 compatibility since it's, you know, metal fuel and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 metal coolant. And so through a lot of the tests, the 1
amazing tests really that were done in the EBR-II 2
reactor, they demonstrated that compatibility with the 3
coolant, and so that benefits us a great deal in the 4
safety area. And then it also behaves differently, 5
you know, with regard to reactivity, feedback effects, 6
and things like that.
7 We've also done something, we've removed 8
a big source of sodium-water interaction. So a lot of 9
historical SFRs have delivered heat from the sodium 10 system directly into a steam generator to go convert 11 that to steam and make electricity. And one of the 12 major challenges to safety with that was this 13 energetic reaction between liquid sodium and water.
14 And so by imposing the energy island with a molten 15 salt system, we've taken away that safety challenge 16 between sodium and water, so that was a big 17 improvement made just in the architecture of Natrium.
18 And I've also already referred to the 19 large thermal inertia that kind of gives us that 20 simplified response to abnormal events.
21 So if we look over to the right --
22 MR. BLEY: Dennis Bley. I think you 23 mentioned it earlier, but the salt and the sodium are 24 pretty compatible? They don't have any problems when 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 they interact?
1 MR. WILLIAMS: They do, they do. I was 2
going to talk about that a little bit when we get to 3
the salt chemistry slide, but I'll go ahead and --
4 MR. BLEY: That's fine, that's fine. You 5
can wait.
6 MR. WILLIAMS: All right. I'll mention it 7
then. So if we look at the right side of this slide, 8
we've just simplified this to the three fundamental 9
safety functions of control, cool, and contain, and 10 I'll kind of point out the major design features under 11 each one of these. You can kind of see they follow a 12 pattern of having a normal means of performing a 13 safety function, a passive means of the safety 14 function, plus an inherent means of providing the 15 safety function.
16 And so for control, normal control, I 17 think, is quite similar to what other reactors do. We 18 have motor-driven control rods that can control 19 reactivity and control power level during normal 20 operation and during maneuvers. We also have a 21 gravity driven control rod scram similar to other 22 reactors. But when it gets to inherent reactivity 23 control, it gets different because of the design of 24 the metal fuel and the core restraint system, how the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 core is mechanically designed, there is a large amount 1
of inherent negative reactivity feedback fundamentally 2
in the core design, and so what that means is that we 3
can look at these events that we call unprotected 4
transients where, essentially, we postulate that none 5
of the control rods go in by gravity and, in a certain 6
number of anticipated operational occurrence type 7
events, with none of those control rods going in, we 8
can bring the reactor not down to full shutdown but it 9
can sort of self-correct itself down to a low power 10 level and stabilize just using the inherent reactivity 11 feedback. So that is an effect that was demonstrated 12 a lot in EBR-II transients and it's a factor in this 13 core design, as well.
14 MR. BLEY: In terms of, you mentioned 15 about 800 thermal --
16 MR. WILLIAMS: 840.
17 MR. BLEY: -- megawatts. Yes. So from a 18 size standpoint, I am assuming that your team has 19 optimized, this is a lot larger core and power than 20 EBR-II, so you've optimized, I think Levesque is going 21 to explore this in the closed session a little more, 22 it's about as large as you can go, let me say it this 23 way, and still get the leakage you need so that you 24 get that reactivity feedback that you're looking for.
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24 MR. WILLIAMS: I would say, we can get 1
into more specifics in the closed session, but I would 2
say it's not as large as you can go, but it's as large 3
as you can go with current technology without having 4
to take some leaps forward in other areas of the 5
technology, which we do plan to do. So that's a good 6
point.
7 MEMBER MARCH-LEUBA: So what's the 8
enrichment on cycle length?
9 MR. WILLIAMS: Enrichment is HALEU, so 10 we're up to, you know, 19 percent, just under 20.
11 MEMBER MARCH-LEUBA: And the cycle length 12 that you expect?
13 MR. WILLIAMS: So the initial core loads 14 are going to be one-year cycle lengths, and the reason 15 for that is because our initial fuel type that we're 16 starting the reactor with is something that we want to 17 advance using the lead test assembly program. And we 18 want to collect data, so we want to be taking fuel 19 normally at normal intervals out of the reactor and 20 sending it out for testing to form the qualification.
21 MEMBER MARCH-LEUBA: So testing on the 22 full plan power.
23 MR. WILLIAMS: Exactly. So it will take 24 several cycles of testing with the start-up fuel form 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 to then collect enough data to qualify and go to the 1
advanced fuel. And then cycles will expand to two 2
years and a lot less number of assemblies needing four 3
reloads.
4 MEMBER MARCH-LEUBA: And what's your 5
break-in temperature?
6 MR. WILLIAMS: It's a 510-degree C outlet 7
and 360-degree C inlet.
8 MEMBER PETTI: So the startup fuel is 9
uranium-based, but the advanced fuel is the uranium 10 plutonium?
11 MR. WILLIAMS: No, it's all uranium-based.
12 MEMBER PETTI: It's all uranium-based.
13 MR. WILLIAMS: Yes, all uranium-based.
14 Yes, yes. In the closed session, I'll talk about the 15 differences there. Any other questions on control?
16 Okay. So when we look at cooling, again, 17 well, the main thing is to hold on to the coolant, so 18 we don't have any penetrations through the reactor 19 vessel. The only penetrations into the reactor vessel 20 go through the reactor vessel head, so there's really, 21 you know, low probability of causing a loss of coolant 22 accident. I'll also talk a little bit on another 23 slide about how the pressures of the systems are 24 designed to force leaks inward towards the reactor 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
26 vessel, another feature to prevent losses of 1
inventory.
2 And then we have, you know, in addition to 3
the primary heat transport system which, you know, 4
delivers heat throughout the reactor vessel, it's an 5
integrated primary system, so all of the components of 6
the primary system are inside this reactor vessel in 7
the most compact way that we can lay them out. We 8
also have an intermediate system to this that's also 9
a sodium system, so most historical SFRs do have an 10 intermediate sodium system. We do, as well. And this 11 system can remove heat through both, can remove heat 12 forced circulation or through natural circulation.
13 And it has within it a sodium-to-air heat exchanger, 14 so it's a very effective way of removing heat when 15 we're down in the refueling temperatures or modes like 16 that where we can remove heat and totally decouple 17 ourselves from the energy island entirely by isolating 18 ourselves, in fact, and remove heat during refueling 19 through this sodium-to-air heat exchanger.
20 And then in emergency situations where 21 power isn't available --
22 MR. BLEY: I'm sorry. This is Dennis Bley 23 again. Where is that thing located? Is it inside 24 this containment picture you got here?
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27 MR. WILLIAMS: Yes. Let me get to the 1
next slide because I'll show you real clearly where 2
that heat exchanger is located, yes. Just trying to 3
introduce the functions here and then we'll see where 4
the equipment is actually physically laid out as we go 5
through the rest of the presentation.
6 And so, finally, that's a non-safety way 7
of removing heat. When we go down to the safety means 8
of decay heat removal during a design basis accident, 9
we have something called the reactor air cooling 10 system. This is a system that goes back to PRISM 11 design, traces its lineage back there, and its way of 12 removing heat through the outside wall of the vessel 13 using natural draft air cooling. And so it's a system 14 that's pretty robust because it's always on, so we're 15 allowing a certain amount of heat to be lost through 16 the system during normal operation to get the benefit 17 of having this system always on and ready to remove 18 heat in an accident scenario. And, in fact, because 19 it's based on radiation heat transfer and, you know, 20 as a function of temperature to the fourth power, it 21 really doesn't start to kick in until you get a large 22 enough temperature difference going. And so we're 23 able to design this system to have pretty low 24 parasitic heat losses during normal operation and then 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 really crank in when a design basis accident happens.
1 And so that's our really robust way of removing heat 2
during design basis accidents.
3 And then, finally, on containment, one of 4
the key features here is having low-pressure systems 5
throughout the plant. And so there's really a low 6
driving force for leakage going out the reactor 7
vessel. And, of course, design basis accidents have 8
an intact primary system, so we're not dealing with 9
LOCAs and things like that, so it's really a matter of 10 keeping leakage rates as low as we possibly can 11 through the reactor vessel head.
12 One of the key features here with sodium 13 is it's affinity for radionuclides. So the fission 14 products release through failed fuel will have to flow 15 through a large depth of sub-cooled sodium, which will 16 remove the iodine and cesium and other things that 17 will add to our mechanistic source term analysis. And 18 then we have multiple radionuclide retention barriers, 19 so you've got the ability of the fuel matrix itself to 20 hold on to fission products, we've got the intact 21 primary system, we've got the cover gas above the 22 liquid sodium, and we've got several barriers outside 23 of that, physical barriers to prevent, reduce 24 releases, as well.
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29 MEMBER PETTI: So you didn't specifically 1
use the words functional containment as the SECY.
2 MR. WILLIAMS: Yes.
3 MEMBER PETTI: Do you think of it as a 4
functional containment?
5 MR. WILLIAMS: Yes, yes.
6 MEMBER PETTI: Okay. And there's no 7
additional containment building.
8 MR. WILLIAMS: Right, right.
9 MEMBER PETTI: Okay.
10 MR. WILLIAMS: It's formed by multiple 11 structure systems and components, yes.
12 Okay. So if you put all of that together, 13 that's where you get to a pretty simplified response 14 to abnormal events. They really all follow the same 15 pattern. There is a scram set point that gets 16 triggers and a reliable reactor shutdown takes place, 17 and then there's a smooth transition to natural 18 circulation cooling. The reactor air cooling system 19 kicks in. It takes a long time, actually, to heat up 20 all of this liquid and start to kick in the reactor 21 air cooling system, and that starts to remove decay 22 heat, essentially, indefinitely because it's just 23 using air. There's no need to replenish the ultimate 24 heat sink or anything like that. It's a low-pressure 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
30 functional containment that will continue to prevent 1
uncontrolled releases of radionuclides, and there's 2
really no reliance on the energy island for any safety 3
functions. There's actually no reactor protection 4
system parameters on the energy island, so there's 5
nothing out on that side of the plant beyond the 6
nuclear island that could threaten core cooling, so no 7
scrams needed from that standpoint.
8 MEMBER MARCH-LEUBA: What is the red 9
stuff, what's the green stuff, what's the blue stuff?
10 MR. WILLIAMS: Oh, yes, I do have another 11 slide, I keep saying this, that will talk about that.
12 MEMBER MARCH-LEUBA: I don't believe you.
13 (Laughter.)
14 MR. WILLIAMS: You don't believe me?
15 Okay.
16 MEMBER MARCH-LEUBA: But you keep talking 17 about this air cooling, and I'm thinking it's inside 18 of those green things.
19 MR. WILLIAMS: Okay. Yes, let me point 20 out some of this now. The next slide I have talks 21 about the flow path within the reactor vessel, so it 22 will just go a little bit more detail. Okay. So we 23 have, the reactor vessel is laid out, and it actually 24 has multiple --
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31 MEMBER MARCH-LEUBA: The only way to do it 1
is with a mouse, yes.
2 MR. WILLIAMS: Yes, yes, look at that.
3 All right. Okay. So there's actually physical 4
barriers that separate the reactor vessel into two 5
main pools that we call the hot pool or the cold pool.
6 The red region is meant to indicate the hot pool, so 7
that's what you see here from the top surface of the 8
reactor vessel, it goes all the way down through this 9
central region that we call the upper internal 10 spectrum because it --
11 MEMBER MARCH-LEUBA: Is that like a core 12 barrel in the --
13 MR. WILLIAMS: Yes, it's within the core 14 and the core barrel and it's bounded by, even down 15 here, by the core grid plate at the bottom of the 16 core. And then that separates you from the blue 17 region, which we call the cold pool, which goes from 18 the intermediate heat exchanger exit all the way down 19 to the reactor vessel. It's where the pumps take 20 suction from, so the pumps will take suction from the 21 cold pool and deliver it to the bottom of the core.
22 We call that the high-pressure plenum at the bottom of 23 the core, and then flow goes out, that little darker 24 shaded red box is really the heated length region of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 the core. You can see how small that is relative to 1
the overall height of sodium above it.
2 MEMBER MARCH-LEUBA: Give me a reference.
3 How tall is that? Two meters, five meters?
4 MR. WILLIAMS: It's about 1.3 meters, and 5
the overall height of the fuel assemblies are, like, 6
15 feet. I'm mixing units there. So flow comes up 7
through the core and into this hot pool, and then it 8
enters these heat exchangers. There's two of them.
9 They're kidney-shaped heat exchangers you see on here, 10 and the hot pool fluid is taken into these heat 11 exchangers and flows down and discharges into the cold 12 pool that's in blue. The pumps, like I said, take 13 suction from there and return the flow to the core 14 inlet.
15 So there's also free surfaces in the 16 reactor vessel, the pool reactor. So there's a liner 17 region that goes along the outside of the reactor 18 vessel that is connected to the cold pool, so that 19 means the reactor vessel wall is seeing cold pool 20 temperatures and not hot pool temperatures, and it 21 means that there's a level difference there that is 22 equal to the unrecoverable pressure drop through the 23 IHX. So there's a level difference there, and when 24 you shut the pumps off that equalizes, yes.
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33 So those are kind of some of the unique 1
things about the way the reactor vessel is laid out.
2 And in some of our analyses, we also call something a 3
warm pool, but I won't get into that. That's just 4
sort of an intermediate region that goes around the 5
IHX.
6 VICE CHAIR KIRCHNER: The intermediate 7
coolant loop is a salt loop, right?
9 VICE CHAIR KIRCHNER: So you've made the 10 statement that the nuclear island's --
11 MR. BLEY: Use your mike, please, Walt.
12 VICE CHAIR KIRCHNER: Sorry. It's on 13 Dennis. I'll get a little closer to it. That the 14 nuclear island is separated from -- what do you call 15 the balance of the plant here?
16 MR. WILLIAMS: Energy island.
17 VICE CHAIR KIRCHNER: Energy island.
18 MR. WILLIAMS: Yes.
19 VICE CHAIR KIRCHNER: Do you have to worry 20 about freezing your sodium, or is the salt melting 21 point in the energy island at a high enough 22 temperature that you don't have to worry about 23 overcooling transients? That's a way that that non-24 nuclear island could couple in an adverse manner with 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
34 your nuclear island.
1 (Whereupon, the above-entitled matter went 2
off the record at 1:38 p.m. and then went back on the 3
record at 1:38 p.m.)
4 MR. WILLIAMS: Okay. All right. So there 5
was, I think there was kind of two questions in there.
6 One had to do with are you worried about freezing, and 7
the other one about overcooling. And I think we are, 8
of course, looking at overcooling as part of the 9
safety analysis. You always have to do that, so, you 10 know, pump overspeeds and heat exchanger being more 11 effective than you intend them to be or turning on, 12 blowers turning on when you don't expect them to are 13 all accident initiators that we've looked at and don't 14 see anything real problematic there, again, because of 15 all that thermal inertia that we have.
16 Freezing is certainly something that we 17 have to manage and engineer our way through. During 18 normal operation, there's not a challenge to freezing 19 because everything is nice and hot. But if you were 20 shut down for an extended period of time, we have heat 21 tracing on all of the sodium and salt piping to 22 prevent freezing from happening, and it's also there 23 to allow for maintenance because, when you want to do 24 maintenance, you do want to freeze to actually allow 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 you to perform maintenance. So there's a lot of heat 1
tracing in the reactor for that reason. It's pretty 2
simple but --
3 MEMBER HALNON: The energy island, though, 4
if it freezes up, it's just an economic issue, it's 5
not a safety issue, correct?
6 MR. WILLIAMS: Right. It's not because we 7
would run back and decouple, close the isolation 8
valves entirely.
9 MEMBER HALNON: Yes, the inherent self 10 island, self nuclear island.
11 MR. WILLIAMS: Right. Okay. So there was 12 a question on ancestries --
13 MR. BLEY: This is Dennis Bley. I don't 14 know if you guys know it, but the meeting timed out 15 about five minutes ago and we can only listen on the 16 phone now. We can't --
17 CHAIR REMPE: Dennis, Dennis, most people 18 have rejoined. There was something that just happened 19 with the software and we got all kicked off for 20 something, but, anyway, most people have been able to 21 rejoin. And I think the court reporter, the court 22 reporter, just to verify, you can hear, right? Right.
23 Thank you.
24 MR. WILLIAMS: Okay. Yes, so if you look 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
36 back at all the sodium fast reactors that form a basis 1
for Natrium, you can go all the way back to EBR-I, 2
although that was really just a demonstration of 3
breeding and wasn't really sodium cooled, it was 4
sodium potassium cooled. But EBR-II and FFTF are the 5
key ones for us. EBR-II with a lot of the passive and 6
inherent safety tests, the run beyond cladding breach 7
tests. And FFTF certainly in terms of all the 8
operating experience with sodium systems and also the 9
tests that they were performing on metal fuel and 10 scram to natural circulation tests. And then even 11 TREAT, the transient reactor test facility, is a 12 source of information on severe accidents with metal 13 fuel.
14 So those are the main areas to look at for 15 the ancestry for Natrium. And like was said earlier, 16 it's a combination of the PRISM and the traveling wave 17 reactors, and I also have to mention Clinch River 18 because the Clinch River breeder reactor had certain 19 parts of its design that are beneficial to us, as 20 well.
21 All right. We spent a lot of time on that 22 slide, but there's a lot on there, so I think that was 23 good. So I think I've talked about most of this, you 24 know. Key differences from light water reactors, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 we've kind of already discussed that. The systems are 1
very compact. You'll see that on the next few slides 2
visually. The low-pressure systems, the efficient 3
heat transfer just because you've got metal fuel, 4
metal coolant, a metal reactor vessel. It's metal, 5
metal, metal, so it's very efficient if you draw the 6
thermal resistance diagram. Pool design with a large 7
coolant inventory is a big factor in safety.
8 Modularity, we've used modular construction wherever 9
we can. It's not a factory-built reactor vessel, but 10 we still use modularity in a lot of the design of the 11 plant and it's architected to really separate safety 12 and non-safety with special treatment systems away 13 from non-safety systems.
14 And also that energy island, nuclear 15 island separation allows for parallel construction 16 techniques, and the extra inherent and passive safety 17 allows us to do a reduced emergency planning zone.
18 MEMBER PETTI: So is it a goal for the EPZ 19 to be the site boundary?
20 MR. WILLIAMS: Yes. All right. So this 21 slide gives you a good view visually of the layout.
22 This is pretty true to what this area in Wyoming 23 actually looks like, so pretty nice view of the plant.
24 And you can see in the center of the slide, I'll use 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
38 my mouse here again, we have the reactor, the three 1
main nuclear island buildings. There's the reactor 2
building in the middle, there's the reactor auxiliary 3
building just next to it, and the fuel handling 4
building just next to that.
5 So those three buildings form the central 6
point of the nuclear island, and the heat essentially 7
comes from below grade in the reactor building out of 8
the reactor vessel through those intermediate heat 9
exchangers that we saw in the other slide, enters the 10 intermediate heat transport system, which is mostly 11 located in the reactor auxiliary building, so the 12 intermediate sodium goes through these sodium to salt 13 heat exchangers and delivers the heat to the salt 14 piping, which leaves the reactor auxiliary building 15 where my mouse is going all the way over to the 16 thermal energy storage tanks, which are over here on 17 the energy island. The thermal energy storage tanks, 18 well, the hot tank essentially pumps salt into the 19 salt-to-water steam generators that are in the steam 20 generator building. Those will generate super heated 21 steam to go to the turbine.
22 And everything that you see beyond the 23 energy storage tanks are sized for 500 megawatts 24 electric, even though the nuclear island is generating 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 345 megawatts electric steady state. So that allows 1
us to ramp up to 500 megawatts electric for about five 2
hours if the grid demands it, but it also allows us to 3
go down to about 100 megawatts electric if the grid 4
demands it and the renewables are working the way 5
they're supposed to and the sun is shining and the 6
wind is blowing, you may want to reduce that heat and 7
save it up for later. So that's the magic of the 8
energy island and the ability to load follow with that 9
while keeping the nuclear island constant.
10 MEMBER HALNON: A couple of things jump 11 out. First of all, you may have your reasons, but I 12 would put the control building more centralized on the 13 other side of the plant just so you can get to energy 14 island quicker. But, nevertheless, if you had a two 15 unit or three or four multiple unit site, what 16 synergies would you have with this single unit besides 17 the switch yard?
18 MR. WILLIAMS: Yes, exactly. So the dual 19 unit site would essentially share the fuel handling 20 building, and so, in fact, one of our requirements is 21 to design the fuel handling building so that a later 22 second unit could be added at a later time.
23 MEMBER HALNON: Just on the other side.
24 MR. WILLIAMS: Just on the other side.
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40 Exactly, yes. And they can also share the energy 1
island. Our site is Wyoming is actually sized to be 2
able to add an additional pair of energy storage 3
tanks.
4 So that's the other thing you can do is 5
the energy island is very modular. There's a lot of 6
optionality on the energy island, depending on the 7
grid and where you're locating the plant, to have 8
multiple salt tank pairs, have a different sized 9
turbine, have multiple turbines. There's lots of 10 things you can do with that to meet whatever that grid 11 demands.
12 MEMBER HALNON: So just back to my first 13 comment, is there a reason why the control building is 14 not centralized? Just as an ex-operator, I would want 15 to be able to get to the energy island rather quickly.
16 MR. WILLIAMS: I think that's a good 17 comment. I mean, I know there were a lot of factors 18 into placing that. I don't recall exactly. I mean, 19 it was safety of providing the cabling that goes from 20 the control building into the reactor building. It 21 also houses the reactor protection system equipment in 22 there, so I know there were a lot of factors in that, 23 but I don't recall exactly why it's there.
24 MR. WERNER: I could maybe help a little 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 bit. This is Mark Werner from TerraPower. One thing 1
to note is that the operators in our control room 2
don't have any safety-related actions, so they don't 3
need to be close to anything to do anything. We do 4
have control, something like a control room in the 5
energy island to provide local control of, you know, 6
near the turbine facility, so there will be people in 7
that area that will provide local control. We are a 8
dispatchable power type of plant, so there's a little 9
bit less to do over there.
10 But one of the things we did try to do on 11 this plant, as Eric mentioned, is a little bit more of 12 a distributed architecture. From a construction 13 standpoint, that control room can be constructed with 14 a work front at the same time as the other facilities.
15 And we see a lot of value in shortening our overall 16 construction duration from a delivery standpoint. So 17 by separating all these buildings, providing a little 18 bit of space, we can have four or five separate crews 19 working on facilities at the same time. If you put 20 them all on top of each other, then you have to wait 21 for the mechanical team to finish, the civil team to 22 finish, before you can start laying down your 23 electrical work. By separating them, they can be 24 working at the same time, shortening construction.
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42 MEMBER HALNON: Thanks. Acreage-wise, 1
what is that about, what we're looking at? Five, five 2
acres?
3 MR. WERNER: The overall site is about 50 4
acres. The nuclear island itself is about 16 acres, 5
and it's still fairly distributed, you know. Eric 6
mentioned kind of the flexibility of our layout and 7
flexibility of our energy island, you know. How we 8
orient all of these facilities is really not coupled 9
to our safety case at all, and so if we had to deploy 10 a site not like Wyoming and space was a real concern, 11 we could make things closer.
12 VICE CHAIR KIRCHNER: You want to orient 13 your turbine building in a preferential way.
14 MR. WERNER: Sure. Right. And to be 15 clear, there are spacing requirements and location 16 requirements. I'm just noting that, you know, if we 17 go to a site that was narrower or longer, because 18 we're decoupled and that salt pipe doesn't lose a lot 19 of heat, isn't that expensive in the grand scheme of 20 things, we can reorient our two sites to kind of fit 21 the local geography pretty well.
22 VICE CHAIR KIRCHNER: Dennis, I see your 23 hand up. Go ahead.
24 MR. BLEY: Yes, I had a couple of things.
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43 One is I kind of agree with Greg. I don't know how 1
long the distance is from the control building over to 2
the energy island, but having a lot of separation from 3
the operators could be troublesome at some times. I 4
understand your construction ideas, too.
5 The idea is, I really like this. This is 6
pretty clever. The other side of the story, if the 7
grid load drops off or the grid trips, it would appear 8
you don't need to scram, you can keep pumping heat 9
over into those energy storage tanks. How long can 10 you do that before you got to ramp down power or trip 11 the reactor?
12 MR. WILLIAMS: Yes, it's essentially 13 another, it's just five hours in both directions.
14 MR. BLEY: Either way. Okay.
15 MR. WILLIAMS: Yes.
16 MR. BLEY: And you run into some kind of 17 mechanical limits on the temperature if you --
18 MR. WILLIAMS: Right. We would be looking 19 at the tank levels, and the tank levels would get to 20 a minimum or maximum --
21 MR. BLEY: Sure.
22 MR. WILLIAMS: Yes.
23 MR. BLEY: Okay. Oh, by the way, please 24 keep talking into your microphones. When you turn 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 your head away, we can't hear you out here.
1 MR. WILLIAMS: Oh, okay. Thanks.
2 MEMBER MARCH-LEUBA: We're having problems 3
today that we typically don't have. On the storage 4
tanks, do you increase the temperature of the salt or 5
do you melt more salt? You said something about 6
level.
7 MR. WILLIAMS: Yes, we don't melt more 8
salt. The salt tank levels change, so, if you are 9
ramping down electrical output, you accumulate hot 10 salt in the hot salt tank, so the level goes up.
11 MEMBER MARCH-LEUBA:
So it's just 12 expansion.
13 MR. WILLIAMS: Yes.
14 MEMBER MARCH-LEUBA: So you're changing 15 the temperature of the salt, not the amount of salt.
16 MR. WILLIAMS: It's the amount of salt in 17 that hot salt tank, yes. It will go up. You will 18 accumulate hot salt if you're throttling back the hot 19 salt --
20 MEMBER MARCH-LEUBA: So what is it the 21 cold salt?
22 MR. WILLIAMS: And the cold salt is in the 23 cold salt tank --
24 MEMBER MARCH-LEUBA: Oh, so you have two 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
45 tanks, one is cold, one is hot.
1 MR. WILLIAMS: Yes.
2 MEMBER MARCH-LEUBA: So you are, it's not 3
that you're melting salt.
4 MR. WILLIAMS: Right.
5 MEMBER MARCH-LEUBA: You're transferring.
6 MR. WILLIAMS: Moving it, yes, yes. It's 7
a constant volume of salt.
8 MR. LEVESQUE: It just, you know, it 9
transfers from one tank to the other as we turn hot 10 salt into cold salt.
11 MR. WILLIAMS: Dr. Rempe, what do we do 12 when we get to 2:00? Do you want me to keep going?
13 VICE CHAIR KIRCHNER: We'll stop where we 14 are, and then we will have people who are joining us 15 and are okay to join us will come in on a different 16 Teams connection --
17 MR. WILLIAMS: Okay.
18 VICE CHAIR KIRCHNER: -- that's been 19 limited. So we'll need to take a break, and we'll 20 need to just make sure everyone in the room, I see 21 you've handed out already proprietary information. I 22 hope you've got a handle on it because the room has 23 been open, but we'll close the room, and then we'll go 24 into the closed session.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
46 So if you're at that juncture, this would 1
probably be a good time for us to take a 10- to 15-2 minute break and reset our Teams connection.
3 MEMBER MARCH-LEUBA: It doesn't need to be 4
exactly at 2:00. It can be at 2:30.
5 CHAIR REMPE: I hadn't heard it had to be 6
exactly.
7 MR. MOORE: This is Scott Moore, the 8
executive director. It does not have to be precisely 9
at that time. The meeting in Teams will remain open.
10 VICE CHAIR KIRCHNER: Okay.
11 CHAIR REMPE: But when you want to go to 12 the closed meeting, it doesn't happen quickly because 13 we've got to check the room and a break is good idea.
14 MR. WILLIAMS: Okay. All right. So this 15 was the vertical cut-through, those central three 16 nuclear island buildings, and so right away you can 17 kind of see that we have the, like I said, the safety 18 related and non-safety with special treatment systems 19 below grade. All the radiological systems are 20 essentially below grade in the reactor. That includes 21 the reactor vessel, it includes the piping of the 22 intermediate heat transport system just above the 23 reactor vessel. And this area that you see, this 24 rectangle above the reactor vessel is what we call the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 head access area, so there's a lot of equipment, very 1
important equipment in that region.
2 Over below grade in the reactor auxiliary 3
building, we've got radiological systems there for 4
processing both the liquid sodium from the reactor 5
vessel and the gaseous, the gases from the reactor 6
vessel, so those are all below grade. And then above 7
grade, you've got the sodium-to-salt heat exchangers, 8
essentially, and the intermediate sodium pumps that 9
make up the intermediate heat transport system.
10 You can also see on the right the fuel 11 handling building, spent fuel pool below grade. And 12 then in both the reactor building and fuel handling 13 buildings, you can kind of see they're very open 14 structures, so the refueling maneuvers can happen 15 between these two buildings. It's not that we've 16 forgotten to add systems in these buildings. They 17 really are open. And you can see that, because the 18 radiological systems are below grade in this nuclear 19 seismic grade concrete structures, we can afford to 20 take a different approach above grade, and we have, 21 essentially, steel structures above grade. So that's 22 where a lot of the modularity comes in, too. These 23 steel structures, you know, can be handled much 24 differently and economically.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
48 I had a question earlier about where was 1
the intermediate air cooling heat exchanger, and this 2
was the slide that shows that very clearly. Over here 3
on the side of the reactor auxiliary building, outside 4
of the reactor auxiliary building, and so that is a 5
sodium-to-air heat exchanger connected to the 6
intermediate heat transport loop and it allows through 7
both forced cooling and natural draft cooling to take 8
heat out of the reactor vessel and deliver it to the 9
air without using the safety related system. And so 10 that's the normal decay heat removal system that we 11 would use in refueling modes and hot standby and so 12 forth.
13 You can also see the ducts for the reactor 14 air cooling system here, and so the air is flowing 15 down and around the reactor vessel and it's coming in 16 through ducts, and these lower ducts is where it's 17 coming in and it's coming out through the taller 18 ducts. And there's four inlets and four outlets to 19 that system.
20 VICE CHAIR KIRCHNER: What parts of the 21 reactor building do you have to harden against natural 22 hazards in particular?
23 MR. WILLIAMS: Yes, that would be this 24 level at grade that protects the --
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49 VICE CHAIR KIRCHNER: So your reactor air-1 cooling ducts, if they were severed for whatever 2
reason, the system would still function?
3 MR. WILLIAMS: Right. We are looking at 4
that, and they're set up on different sides of the 5
reactor building with enough separation that we can 6
deal with some of those hazards.
7 VICE CHAIR KIRCHNER: One of the 8
vulnerabilities that was, I think, there in the PRISM 9
design was that they had that functional heat 10 exchanger above grade, potentially exposed to hazards.
11 So, basically, if these ducts sever, you'll still get 12 a natural draft through the reactor cavity to take out 13 the decay heat.
14 MR. WILLIAMS: They are very robust with 15 regards to, you know, if you sever two of them and you 16 still have two functioning or if you increase pressure 17 drop in these ducts. A lot of the testing that was 18 done on the system was done at Argonne National Lab 19 and showed that it really wasn't as sensitive to 20 pressure drop as you would think. It's more radiation 21 heat transfer limited, and so it's a very unique kind 22 of robust system. But there's all kinds of external 23 hazards also that can degrade the system that we look 24 at, you know, debris getting entrained into the ducts, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
50 and there's all sorts of degradation mechanisms that 1
were taken into account to --
2 VICE CHAIR KIRCHNER: Dennis, do you have 3
your hand up again?
4 MR. BLEY: I do. This is Dennis Bley.
5 VICE CHAIR KIRCHNER: Okay. Go ahead.
6 MR. BLEY: It looks like if you've 7
hardened the aux building, things are probably pretty 8
good, except airplane crash, I assume, could take out 9
the intermediate air cooling system; is that right?
10 MR. WILLIAMS: The intermediate, there's 11 two intermediate air cooling heat exchangers on sides 12 of the reactor aux building, so I'm not, I can't 13 answer that exactly whether it can take out both of 14 them or not; I'd have to go back and look. But those 15 aren't safety-related systems either.
16 MEMBER MARCH-LEUBA: Your safety-related 17 air cooling is inside of the reactor cavity.
18 MR. WILLIAMS: Right.
19 MEMBER MARCH-LEUBA: And where does the 20 heat go?
21 MR. WILLIAMS: So the heat --
22 MR. BLEY: Yes, you were going to show us 23 that.
24 MR. WILLIAMS: Yes. So the heat, it comes 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 in through the reactor air-cooling ducts that I'm 1
pointing to with my mouse. There's two of them shown 2
here. There would be two, a mirror image, on the 3
other side. So four air intake ducts.
4 MEMBER MARCH-LEUBA: That reactor cavity 5
has a couple of holes that are connected with the 6
atmosphere.
7 MR. WILLIAMS: Absolutely. It's all 8
connected to the atmosphere. And so it comes down --
9 MEMBER MARCH-LEUBA: The one, it looks 10 like a containment fluoride. It's not.
11 MR. WILLIAMS: Right. Well, yes. So let 12 me try to --
13 MEMBER MARCH-LEUBA: You have a way for --
14 I couldn't figure out where the heat was going.
15 MR. WILLIAMS: Okay. We do have more 16 sketches of this in the closed session that show the 17 details.
18 MEMBER PETTI: Do you use a seismic 19 isolator on the --
20 MR. WILLIAMS: We don't seismically 21 isolate the plant at large, but we have seismic 22 isolation on the reactor vessel.
23 MEMBER PETTI: The reactor vessel. Good.
24 MR. BLEY: I guess, before you leave this, 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
52 I'm still thinking about an aircraft crash, and that 1
reactor building up there, that's a steel building it 2
looks like. If that gets taken out, it seems like 3
those air ducts and things could all get crushed up 4
and wouldn't work anymore.
5 MR. WILLIAMS: Yes, we'll have to show 6
that there's enough air cooling even with an airplane 7
crash.
8 VICE CHAIR KIRCHNER: But you wouldn't 9
want collateral damage if you have an aircraft event 10 hitting that sodium intermediate loop and then 11 spreading sodium into the reactor building cavity.
12 MEMBER MARCH-LEUBA: On a rainy day.
13 MR. WILLIAMS: Yes, on a rainy day.
14 MR. BLEY: Okay. We'll see that kind of 15 detail later, but I'm curious about how you deal with 16 that.
17 MR. WILLIAMS: Oh, Tara did mention 18 something to me real quick. There's been an evolution 19 to this design, and the reactor, it looks like the 20 reactor air-cooling ducts are integral with the 21 reactor building, but they're actually separate from 22 the building. So the reactor air-cooling ducts are 23 safety-related ducts, and the reactor building is not.
24 So those are separate.
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53 MR. BLEY: Okay. There's a great big 1
crane up there that could come down with the building, 2
though, it looks like. Pretty heavy.
3 MR. WILLIAMS: We'll jump into 3D model in 4
the closed session, and we can look at this in detail 5
of the current design. This is a pre-conceptual 6
image.
7 Okay. So I'll go through this one a 8
little quickly. We talked a bit about the energy 9
island thermal storage. The great part about this 10 side of the plant is that all of this technology is 11 readily available from the concentrated solar plant 12 industry, and so that was one of the benefits of why 13 we chose it. We even chose the same salt composition 14 that is typically used in concentrated solar plants to 15 make their, you know, good access to the OE database 16 of all of that work.
17 So when you look at this, there's lots of 18 different options, from the number of tanks, the 19 number of steam generator trains, and what sized 20 turbine you need for the grid where you want to put 21 this plant. The picture on this slide on the left 22 kind of shows you a top-down view of a concentrated 23 solar plant, and you can see how many salt tank pairs 24 they have. So they've definitely chosen to go with a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 higher degree of thermal energy storage for this 1
location, and, like we said earlier, that's something 2
that you can do with the Natrium design, as well.
3 All right. So a few things to just talk 4
about the benefits of sodium and then the benefits of 5
salt. When we look at sodium coolant, and there's 6
been a lot of work with liquid sodium and liquid 7
sodium reactors in the United States, the one that's 8
really not, that's not really mentioned on this slide 9
is that sodium is chosen for its neutronic properties 10 because it doesn't moderate neutrons. So if it were 11 to have a higher degree of moderation, it would soften 12 the spectrum, and we may not have a fast reactor 13 anymore. So it's chosen for the reasons of wanting 14 the sodium fast reactor, and then the higher heat 15 capacity that it provides allows us to have reasonable 16 sized pumps. The high heat transfer is really the 17 most important thing. It allows the core to be 18 reasonable sized, and it allows that great decay heat 19 removal that I talked about and allows us to use these 20 sodium-to-air heat exchangers that are so effective 21 and so valuable to us in decay heat removal.
22 It's also got a good range for where we 23 want to operate, so the high boiling point around 883 24 degrees C and a melting point around 98 degrees C, and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 all of that is at atmospheric pressure, so that means 1
we can have a high temperature heat output coming from 2
the reactor vessel, which allows us to have high 3
thermal efficiency without having to pressurize it.
4 So that's the key to being at atmospheric pressure.
5 And then the density of sodium is 6
remarkably similar to water, which allows us to do a 7
lot of testing, do the scaling of a test for sodium in 8
a water system and actually get that to work, and so 9
reducing the costing of testing and increasing the 10 value of testing with water. Lack of corrosion, so 11 there's been demonstrations of very low corrosion in 12 sodium systems over time. We learned a lot of that 13 from FFTF. Limited auxiliaries, like I mentioned, a 14 couple of systems to remove non-condensable gases and 15 impurities from the sodium liquid and the gas; but 16 other than that, there's not a whole lot of 17 conditioning that you have to do.
18 And then there's a large sodium inventory, 19 so 800 cubic meters of sodium in the reactor to absorb 20 all of that decay heat removal when you need it. And 21 the graphs here show you the differences between, 22 like, the peak, the cladding temperature and the peak 23 central temperature of the fuel between an oxide fuel 24 and metallic fuel. And so because metallic fuel can 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
56 remove heat so effectively from the fuel itself, you 1
actually have a pretty close delta between the fuel 2
temperature and the cladding temperature shown on the 3
right, whereas an oxide fuel you don't. You have a 4
higher center line fuel temperature that is sometimes 5
above the boiling temperature of the coolant. And in 6
sodium reactor oxide fuel reactors, they have that.
7 So the normal operating fuel temperature was higher 8
than the boiling temperature, but in Natrium it's 9
several hundred degrees below the boiling temperature 10 because of metal fuel. So those were some of the 11 great benefits of using sodium.
12 And then this next slide just shows some 13 pictures. If you come out and visit us in our lab, 14 we'll let you take a look at these up close but not 15 too close, and we even let people cut into solid 16 sodium to kind of get a sense of what it feels like.
17 It kind of feels like cutting through butter taken out 18 of the refrigerator or something. But you can see 19 it's opaque, so that's kind of an important property 20 that we have to be concerned with in terms of 21 refueling. A lot of times, it builds up an oxide 22 layer on the top. It has an auto-ignition temperature 23 that varies quite widely between about 100 degrees 24 Celsius and 400 degrees Celsius. So, you know, you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
57 could have sodium sitting there below its auto-1 ignition temperature, but, of course, at reactor 2
temperatures, it's going to be above that.
3 And on the solid side, you can kind of see 4
it's built up this oxide film. It's kind of grayish 5
or even pinkish sometimes.
6 And so one of the things with sodium is it 7
does react energetically with both water and air, and 8
so, for our sodium systems, we have to prevent and be 9
prepared to detect and mitigate any sodium fires. So 10 one of the downsides to using sodium is, in addition 11 to the heat tracing we mentioned earlier, is providing 12 all that sodium fire protection.
13 And so, you know, it's a kind of 14 engineered safety system that has to exist in the 15 plant. But the good news is it's all at atmospheric 16 pressure, so, when we have leaks, they're not, you 17 know, you sometimes see in the literature spray fires 18 and pool fires and things like that. What we 19 typically call it is a drift fire. It would be like 20 a slow leak at a flange or something like that.
21 Typically, these leaks are, you know, they don't get 22 through the guard piping and inerted space that we 23 have the pipe in, and so they're detected by the 24 sodium leak detectors, which then allow us to go in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 there and perform the maintenance --
1 MEMBER PETTI: So all the pipes are 2
guarded.
3 MR. WILLIAMS: All the pipes have some 4
kind of, and we have a slide on this, some kind of 5
leak protection to them. There's all different ways 6
depending on where the pipe is.
7 MEMBER PETTI: The opacity I was aware of, 8
but the program inspection and, you know, how you see 9
into this code requirements. I know, obviously, some 10 of the components are above the level, but you 11 probably still have a need to get down lower with the 12 sodium. Has the technology evolved that you've got a 13 solution?
14 MR. WILLIAMS: We're still looking at, you 15 know, technologies for under sodium viewing, so that 16 is certainly a part of a project is to look into 17 those. But we're also looking at, you know, risk-18 informed inspections and different ways of inspecting 19 equipment that is below the level of sodium, so it's 20 kind of a combination of all of them. But for the 21 refueling, you know, we will have discrimination 22 features on the top of the fuel assemblies that will 23 allow all of that to occur, even under the opaque 24 sodium. And fuel assemblies are inspected when they 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
59 come out and they get cleaned and all of that, as 1
well.
2 MEMBER BALLINGER: This is Ron Ballinger.
3 Are you in any way connected with or taking advantage 4
of the French side, especially with respect to 5
inspection, refueling, those kinds of things with 6
Phoenix and Super Phoenix?
7 MR. WILLIAMS: We're still working on that 8
front. We'd like to be more connected. I don't think 9
we are -- but we are talking right now.
10 MR. WERNER: I think we're sending a team 11 over, I think we're sending a team over, like, next 12 month to collaborate on methods and operations, yes.
13 And this is Mark Werner.
14 VICE CHAIR KIRCHNER: Also look at 15 construction, too, because if you study the Super 16 Phoenix experience.
17 MEMBER BALLINGER: They had some infamous 18 issues, like not being able to find the fuel elements.
19 MR. WILLIAMS: Okay. That would be 20 infamous. All right. So another couple of slides 21 just on salt. So we are using nitrate salt, you know, 22 60-percent sodium nitrate, 40-percent potassium 23 nitrate. This is the typical composition used in the 24 solar industry. Molten salt inventory is quite large 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
60 with those large thermal storage tanks, 30,000 tons.
1 So there's quite a bit of thermal energy storage 2
capacity that we have over there, and, of course, 3
anything happening on the other side of those tanks 4
with the steam generators, the turbine, or all of that 5
equipment, just can't make its way back to the nuclear 6
island very quickly.
7 So there's lots of benefits of using this 8
molten salt. It actually does have pretty good long 9
life with really low performance degradation, and it's 10 really good in the range of temperatures that we want.
11 We can use carbon steel up until about 400 degrees C 12 and stainless steel up to about 600 degrees C, so the 13 cold tank side is typically carbon steel and the hot 14 salt tank side is typically stainless steel. So a 15 really good temperature range for us and a really high 16 degree of efficiency in terms of retaining the heat.
17 And so a lot of people ask us, you know, are you 18 losing a lot of heat from the tanks that are sitting 19 there, you know, during outages and things like that 20 and not being used, and the answer is no. Very little 21 heat is lost through them.
22 And so, you know, a really common use in 23 the industry. I don't think it's been used in a 24 nuclear facility, so, you know, we are taking a look 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
61 at, you know, the operating experience and history 1
with designing these tanks in the industry and making 2
sure that they are designed to the right level of 3
reliability for us.
4 MEMBER BALLINGER: Do you have any 5
flammability issues with this salt mixture?
6 MR. WILLIAMS: No. So salt is not 7
flammable with air, so when salt leaks, you know, you 8
go and clean it up with a shovel. It solidifies on 9
the outside, and there's no danger with that.
10 I was going to mention the salt and the 11 sodium is an exothermic reaction. It's relatively 12 mild. It's nothing like sodium and water, but it is 13 one of the technology factors that we are doing quite 14 a bit of testing on right now in our lab in 15 Washington. And so we have a couple of systems, a 16 couple of test loops running right now with sodium-17 salt interaction tests, and what the data are really 18 showing us is that when you have, in our case you 19 would have salt leaking into a sodium system, it does 20 generate some energy, but it is relatively mild. And 21 because the sodium is so thermally conductive, the 22 heat dissipates very quickly in the sodium system, and 23 you can detect it by the presence of, I think it's 24 hydrogen and nitrogen that are generated from this 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
62 reaction, and those are gasses that we're already 1
looking at in the intermediate heat transport system.
2 So you can detect a leak when it happens, and you can 3
shut down and perform maintenance. You can train the 4
heat exchanger and perform maintenance on that. But 5
we're still continuing to test that.
6 Yes, go ahead.
7 VICE CHAIR KIRCHNER: Go ahead, Dennis.
8 MR. BLEY: You said hydrogen. This is 9
Dennis Bley. I don't see any hydrogen.
10 MR. WILLIAMS: It's the -- oh, yes, you're 11 right. Yes, it must have been nitrogen. Sorry.
12 You're right. There's no hydrogen in there. Thank 13 you for correcting me on that.
14 MEMBER PETTI: And there's no chemical 15 issues in terms of the vapor pressure that workers 16 need protection, besides the temperature of the salt, 17 in terms of maintenance.
18 MR. WILLIAMS: Yes, I don't, I'm not aware 19 of any.
20 MEMBER BALLINGER: There was an incident 21 at Cadarache where they were cleaning a tank which 22 formerly contained sodium and had residue in it, and 23 something happened. Now, I'm a little fuzzy on what 24 they were using, whether it was a cleaning fluid or 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
63 whatever it was, reacted with the residual sodium, and 1
they killed a few people.
2 MR. WILLIAMS: Okay.
3 MEMBER BALLINGER: It was an enclosed 4
tank, and it was a bad hair day.
5 MR. WILLIAMS: Yes.
6 MEMBER BALLINGER: So there are those 7
issues.
8 MR. WILLIAMS: Yes. I think for sodium 9
you have to take a lot more precautions than with the 10 salt.
11 MEMBER BALLINGER:
Certainly with 12 cleaning, with what they were doing.
13 MR. WILLIAMS: Right, right. Good 14 comment.
15 So salt in the molten state, you know, 16 it's the white substance on the left as a solid and a 17 clear liquid as you can see on the right. And so, you 18 know, leaks are, again, something that we have to 19 consider and look at, but nothing like sodium leaks.
20 So we'll definitely be looking at that and having 21 plans for maintenance and recovery after leaks.
22 MR. WERNER: Hey, Eric, this is Mark 23 Werner, just jumping in. If you go back a slide, just 24 to maybe add a little more context to this. The image 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
64 on the left is actually salt that we leaked out into 1
a pan and just collected in a jar. And so when we're 2
worried about leaks, it just solidifies to look like 3
what's on the left. And on the right, you know, 4
that's molten salt, but it's in an air environment.
5 It's not in a glove box or anything, so this kind of 6
image shows how compatible this working fluid is with 7
the environment.
8 MR. WILLIAMS: Thanks, Mark. All right.
9 So when you take a look at the safety systems, we 10 called them, you know, compact and robust for the 11 Natrium reactor. So if you just look at comparison to 12 a light water reactor where you're, you know, your 13 safety systems are daisy-chaining multiple systems to 14 take the heat out of the reactor and take it to its 15 ultimate heat sink, it just generates a lot of, we 16 call it nuclear sprawl sometimes that makes a large 17 fraction of the plant become safety related as a 18 consequence. And so by having these reactor air-19 cooling ducts removing heat using air, you have a more 20 compact system that doesn't result in that same 21 sprawl. But having the integrated primary system all 22 inside the reactor vessel, we've eliminated a lot of 23 Section III pipe welds.
24 Being at atmospheric pressure we've 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 mentioned is a big feature here. And just that 1
unlimited supply of the ultimate heat sink, very 2
important for long-term cooling. Being fully passive, 3
having a system actually always in operation, and just 4
having a very rugged system. So there's a lot of good 5
benefits to the technology in that area.
6 All right. So I'm getting close to the 7
end here.
8 MR. BLEY: Dennis Bley again. I almost 9
forgot my question when the picture went away. How 10 hot does the sodium have to get -- I could have done 11 the calculation, I guess, but I didn't -- before this 12 transfer, heat transfer mechanism becomes real 13 effective? It's got to go up a hell of a lot, I would 14 think.
15 MR. WILLIAMS: Yes, yes, it's probably 16 going to go up, you know, 50 to 100 degrees, I would 17 guess. You know, it's always going to be ramping up 18 as temperature is going up. It just takes a long time 19 for that much inventory to raise its temperature. But 20 the initial, you know, following an accident with only 21 reactor air cooling available and nothing else, you 22 know, you would see a lot of the decay heat turning 23 into sensible heat, increasing the temperature of that 24 liquid pool. And then, you know, over the course of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
66 a day or more, you would start to see it start to kick 1
in more.
2 MR. BLEY: Okay. That gives me a hint 3
anyway.
4 MR. WILLIAMS: Okay. So this was the one 5
I was promising about the flow path. It shows kind of 6
what I described earlier, but I think it gives you 7
just a little bit of a clear image between the red hot 8
pool and the blue cold pool. And the arrows show you 9
the flow path through the heat exchangers and the 10 pumps.
11 So a couple of features I'll just point to 12 that we haven't talked about yet is there is a guard 13 vessel outside of the reactor vessel. So one of the 14 ways of preventing, you know, if there was a leak, and 15 this would be a beyond design basis leak from the 16 reactor vessel, we don't want to have interaction with 17 air. So there's a guard vessel surrounding the 18 reactor vessel and an inerted space in between them.
19 And then the gap between that reactor vessel and guard 20 vessel is sized such that, if there was a reactor 21 vessel leak, the total liquid level would still remain 22 above the pumps and the heat exchangers inside the 23 reactor vessel. So even if you had a reactor vessel 24 leak, you'd keep those components covered and you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
67 provide the opportunity to continue to use forced 1
cooling, if you have it, and to still allow natural 2
circulation cooling to occur, of course.
3 MEMBER MARCH-LEUBA: Are the pumps on the 4
outlet of the heat exchanger?
5 MR. WILLIAMS: The pumps take suction from 6
the cold pool and discharge to the core inlet.
7 MEMBER MARCH-LEUBA: Kind of push it. And 8
what drives it through the heat exchanger?
9 MR. WILLIAMS: The heat exchanger takes 10 input from the hot pool at the top and then discharges 11 it into the cold pool.
12 MEMBER MARCH-LEUBA: By gravity?
13 MR. WILLIAMS: Through the pump. The pump 14 is forced cooling up through the core --
15 MEMBER MARCH-LEUBA: So the pump is 16 cooling on it.
17 MR. WILLIAMS: Right.
18 MR. BLEY: Hey, this is Dennis Bley again.
19 On your first picture, the label on the guard vessel 20 said containment. It looks like it is essentially a 21 containment. Why don't you call it one?
22 MR. WILLIAMS: It's a major part of the 23 functional containment strategy. The guard vessel 24 surrounds completely the guard vessel, but it doesn't 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
68 go up around the reactor vessel head. And so you have 1
the seals in the reactor vessel head for the, you 2
know, for the control rod drive mechanisms, and the 3
other penetrations are also part of that functional 4
containment.
5 MR. BLEY: Okay. This picture is not a 6
real picture, so it looks like the vessel head is 7
actually bolted down on to the guard vessel. Is that 8
true, or is that just an artifact?
9 MR. WILLIAMS: Maybe Mark could help me 10 with the interface there between the guard vessel and 11 the reactor vessel.
12 MR. WERNER: Yes. This is Mark Werner 13 from TerraPower. There will be some type of seal 14 between the guard vessel and the reactor vessel. The 15 reactor head will be bolted down on to the reactor 16 support structure, but the guard vessel will be, you 17 know, sealed to the reactor vessel to maintain that 18 boundary.
19 MR. BLEY: Okay.
20 MR. WERNER: We need to keep that inert, 21 as well, and we monitor that space.
22 MR. WILLIAMS: All right. You can see, 23 the only other piece I wanted to point out here was 24 the in-vessel transfer machine shown in blue there and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
69 pointed out with the label. So we are doing refueling 1
inside the reactor vessel. Another kind of new thing, 2
the reactor vessel head doesn't come off. New fuel 3
comes in through an opening in the reactor vessel head 4
and gets manipulated with the in-vessel transfer 5
machine to its location; and, you know, spent fuel is 6
manipulated with the in-vessel transfer machine to 7
remove it from the core. We actually store fuel 8
assemblies after they come out of the core for a 9
period of time in in-vessel storage locations, which 10 you can think of as a couple of rings of empty fuel 11 assembly slots outside the core barrel. So they stay 12 there, they cool off for a cycle, and then the in-13 vessel transfer machine removes them out of the 14 reactor vessel. So that's a unique part of the 15 technology, and so we chose to do it with an in-vessel 16 transfer machine here for Natrium.
17 MR. BLEY: This is Dennis Bley again. In 18 the closed session, are you going to be able to show 19 us how that thing actually works? Does it index 20 around or what's going on there?
21 MR. WILLIAMS: Yes. We've got a movie for 22 you in the closed session to show that.
23 MR. BLEY: Great. I can't wait.
24 MEMBER BALLINGER: This is similar to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 Phoenix.
1 MR. WILLIAMS: I think this is a 2
pantograph, and I'm not sure if they use that at 3
Phoenix. But FFTF used an offset arm type of 4
equipment, and so it was different than this. And so 5
we chose to go with this because we think it will 6
allow us to have more, you know, meet the expectations 7
of commercial refueling times, as opposed to what they 8
did at a test reactor. All right.
9 VICE CHAIR KIRCHNER: So in the closed 10 session, Eric, could you show us where actually in the 11 core the spent fuel is stored temporarily?
12 MR. WILLIAMS: Yes, yes, we have that. So 13 the last slide here before I turn it over to Ryan, 14 we've kind of talked about all of this, but just to 15 wrap up the overall story for heat removal is, you 16 know, you have normal ways of removing heat, which is 17 forced flow through the intermediate heat transport 18 system, using that intermediate air heat exchanger.
19 You also have a passive way of using that same system 20 by not having forced flow and just using natural draft 21 through the intermediate heat transport system, and 22 that intermediate air-cooling heat exchanger has 23 dampers that open and a blower that can also come on 24 and force cool it if you're in the normal mode. And 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
71 then if you're in the passive mode, the dampers just 1
open and it allows natural draft cooling using the 2
same heat exchanger.
3 And then, finally, I call it the inherent 4
decay heat removal, and that's with the reactor air-5 cooling system because it's always on. And that's our 6
safety-related decay heat removal method.
7 So that just kind of summarizes all the 8
ways of getting heat out of the reactor, and now I 9
will turn it over to Ryan who will talk about our 10 licensing strategy.
11 MR. SPRENGEL: Good afternoon. Ryan 12 Sprengel. I'm the Director of Licensing for Natrium.
13 I did check before the meeting. It's been seven and 14 a half years since I've been in front of the 15 Committee, so I see some familiar faces and some new 16 faces. Dare I say, I'm happy to be back in front of 17 you.
18 It's been touched on already, but we are 19 using a Part 50 licensing process. So we'll go 20 through a two-cycle application with our construction 21 permit followed by our operating license application.
22 And highlighted on the slide here, in terms of the 23 approvals for the plant, you know, we'll come back 24 before the ACRS again multiple times. What was 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 mentioned earlier, we'll touch on some of the topical 1
reports, so there's other opportunities to come in 2
front of the ACRS, as well.
3 Part of the reason for the Part 50 process 4
is tied to something that Chris mentioned. You know, 5
we're a part of the ARDP program, so there's an 6
aggressive time frame that we're working under and the 7
two-part Part 50 approach lends itself to starting 8
construction earlier and engaging with the staff 9
earlier to facilitate that time frame.
10 Currently, if we look at this slide, we 11 are to the left of it, so we're kind of off this 12 slide, you know. So we will get our construction 13 permit application submitted and start this flow path.
14 Let's see. Topical reports we'll touch on 15 in a minute, and I'll speak to those time lines 16 specifically.
17 MEMBER MARCH-LEUBA: Just a question. On 18 the construction permit, your intention is to submit 19 it before you really have a finalized design?
20 MR. SPRENGEL: Yes, yes, yes.
21 MEMBER MARCH-LEUBA: It would be kind of 22 what I call aspirational, the statements.
23 MR. SPRENGEL: I don't know if I would 24 consider aspirational, but we will be following on 10 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
73 CFR 50.34, speaking to a preliminary state of the 1
design. As was noted, though, our project --
2 MEMBER MARCH-LEUBA: It will mostly set 3
requirements that you have to meet on the license.
4 MR. SPRENGEL: In some areas, yes. In 5
some areas, there will be things that will be planned 6
out. Some where testing, certainly, will be ongoing.
7 The structures, after we have the CP submitted and 8
reviewed and issued as a construction permit, you 9
know, our project is set up to start construction.
10 MEMBER MARCH-LEUBA: From the typical 11 point of view, when you review a CP, you're very 12 disappointed. There is no information there. So I'll 13 focus myself to adjusting that way.
14 MR. SPRENGEL: Okay.
15 VICE CHAIR KIRCHNER: I think what Jose is 16 referring to is could you give us just maybe some 17 feeling for what the maturity of the design is when 18 you go into the TP stage? Are you ready, at that 19 point, to write procurement specs or you're still, is 20 the design evolving subsequently?
21 MR. WILLIAMS: I'd say it's definitely 22 taking the design down beyond plant functional 23 requirements to system requirements. And then the 24 safety analysis is mature enough to ensure that, you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
74 know, the plant has enough safety margin in it to 1
start construction, so you'll see a lot of the design 2
basis accidents and even beyond design basis 3
accidents, the PRA. Because we're using licensing 4
modernization program, which Ryan is going to talk 5
about here, PRA is a very, you know, factors very 6
prevalently in the design from the beginning, so there 7
will be a lot of that information available. So it 8
doesn't mean we're done with design, but we have to 9
have established the safety margins of the design.
10 MEMBER MARCH-LEUBA: So it won't be a 11 complete PRA, but at the CP stage you expect a decent 12 PRA?
13 MR. WILLIAMS: I do, yes.
14 MEMBER MARCH-LEUBA: Okay. That's good.
15 CHAIR REMPE: You know, I was puzzled with 16 your comment, too, and I guess, you know, we've seen 17 a range of maturities in what we're seeing, and some 18 are very good and, yes, they identify what else is 19 needed, the staff does, and we review it. And then 20 they make changes, and that's wonderful. But it 21 helps, I think, to focus the questions on what you 22 might want to consider that needs to be changed.
23 That's where he's coming from, not that, you know, the 24 way you responded back was a little puzzling to me 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
75 that you were going to change your focus. But, 1
anyway, I think that will be good to see more detail 2
is the bottom line.
3 MR. LEVESQUE: And if I could add, this is 4
Chris Levesque, TerraPower CEO. Just a reminder about 5
how well resourced this project is, I mean, thanks to 6
the Department of Energy and the ARDP grant and thanks 7
to our shareholders. I mean, we have 800 engineers 8
working on the project
- today, getting into 9
considerable design detail, some of which you'll see 10 in the closed session. So there's a lot of rigor and 11 a lot of detail being established today that I think 12 you'll see in the construction permit application.
13 MEMBER MARCH-LEUBA: I'm glad to hear 14 that.
15 MR. SPRENGEL: To date, we have had 16 several engagements with the NRC staff to talk about 17 our PRA. It is integral to our use of LMP. And one 18 of the benefits, we talked about PRISM before and how 19 PRISM has served as, you know, part of the foundation 20 of the Natrium design. PRISM had a very well-21 developed PRA, and we have used that as a really 22 advanced starting point. So we're not starting from 23 nowhere on our PRA, and we've gone further since that 24 time. I do think our PRA is quite advanced at this 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76 point, and, of course, we'll go forward with 1
additional steps of peer reviews, and that's kind of 2
going to be going on in parallel to our CPA. But all 3
that to be said that we are moving forward and we are 4
not trying to just submit the very minimum because, at 5
the end of this, we are planning to construct. We are 6
moving right into activities, and so I think our level 7
of detail should be, hopefully, more advanced to meet 8
the needs.
9 Okay. So licensing modernization project, 10 we are using it. Of course, some of the aspects of 11 it, LB identification, SSC classification, and 12 defense-in-depth, that's just part of the fundamentals 13 of what the LMP process outlines. NEI-18-04, serving 14 the basis of that and the staff's endorsement of it.
15 We're also using Reg Guide 1.232, so recently we've 16 submitted our principal design criteria into the staff 17 January of this year. So that is undergoing staff 18 review at this time.
19 And then at the bottom there, building off 20 of what was mentioned earlier, we're following the 21 ARCAP and TCAP activities for the content of our 22 application. And so NEI-21-07 is one of the kind of 23 foundational documents amongst the many, many draft 24 guidance documents that are out there. We are 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 continuing to follow those and we'll submit under the 1
ARCAP-TCAP kind of structure of the CPA.
2 MEMBER PETTI: So I don't want to put 3
words in your mouth, but it sounds like if Part 53 4
were here today, you've got many of the core elements 5
of Part 53.
6 MR. SPRENGEL: Sure, yes. I think, yes, 7
Part 53 not being here, I guess we haven't gone that 8
far, but yes.
9 MEMBER PETTI: But you've certainly got, 10 I'm going to call it the heart of it.
11 MEMBER MARCH-LEUBA: At ACRS, we fly at a 12 40,000-foot level and trying to see everything. One 13 thing we emphasize a lot is the LB, the licensing 14 basis event selection. It's human nature to start 15 with light water reactor events and scratch out the 16 ones that aren't going to apply to me. What we 17 emphasize a lot is you should not do that. Start with 18 a wide sheet of paper, we hear white sheet of paper a 19 hundred of times while we're here, and think of what 20 could possibly go wrong in your plan that doesn't 21 apply to light waters because it is very human nature 22 to scratch and don't add. So we've been looking into 23 that.
24 MR. SPRENGEL: Thank you for that comment.
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78 We haven't started with light water reactor defense, 1
so it is challenging to, you know, to navigate LMP and 2
understand exactly what's a good LB list. But there's 3
still challenges in following the process because it's 4
brand new, but we're using it.
5 MEMBER MARCH-LEUBA: LMP is great. If you 6
forget an event, it could give you the wrong answer.
7 Bad input, bad answer.
8 MEMBER PETTI: The other thing, to 9
piggyback on that, is you've mentioned safety margin 10 because this is sort of first of a kind. You've built 11 out the sodium system, but there's a lot of newness 12 here. And for us, being at 40,000 feet, you know, to 13 have the safety margin buried in some appendix or some 14 technical document supporting the PSAR doesn't do us, 15 you know, as trying to get the public confidence up, 16 where you guys can come up and say a margin here, this 17 is the margin there, you know. That is a really easy.
18 You'll start to see how our letters on advanced 19 reactors look. They look very different than our 20 letters on water reactors purposely because we 21 recognize the first-of-a-kind nature, and so we're 22 looking at things in an LMP Part 53 sort of framework, 23 you know, top-down because these are so new, you know.
24 So I encourage you to make sure that your 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 safety margins are clearly articulated when you come 1
to us. It makes our job easier to represent you guys 2
properly.
3 MR. SPRENGEL: Thank you. Looking at our 4
topical reports that we submitted, so there's five 5
listed here.
Our quality assurance program 6
description has already been approved. We have a 7
revision that's going through staff review currently 8
in the final stages. I guess, actually, we have the 9
draft SEN in hand right now.
10 George mentioned earlier the energy island 11 interface. That's one that has gone through a staff 12 audit and completed that audit, and so, you know, 13 likely, that would be the next one that would be up 14 for scheduling effort to get in front of the ACRS 15 subcommittee for the Natrium project. I mentioned the 16 principal design criteria here, that was submitted in 17 January of this year. And then the other two are more 18 recently submitted in the last few months, our fuel 19 and control assembly qualification. That's building 20 off of some additional work that we did with the staff 21 and with the Department of Energy on generic advanced 22 fuel qualification, so we used that --
23 MEMBER PETTI: The NUREG that the staff 24 produced, advanced fuel qualification?
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80 MR. SPRENGEL: No. In this case, the 1
delivery was actually to the Department of Energy.
2 MEMBER PETTI: Oh, okay. Because you know 3
there's a NUREG-22, I can't remember the number.
4 MR. SPRENGEL: Yes, yes. Those are, yes, 5
they're interlinked in terms of how they were 6
developed absolutely. The topical report we submitted 7
was specific for TerraPower's fuel and control 8
assembly qualification and also touches on things that 9
Eric mentioned that are kind of next generation of 10 fuel that we have on our radar and is important to our 11 overall project.
12 And then our emergency planning zone 13 methodology, which is -- one part we'll see on the 14 upcoming reports, as well. There's other pieces that 15 really fit into that EPZ methodology.
16 MR. BLEY: Before you leave this, Dennis 17 Bley again. I assume you've done something like a 18 PIRT and identified knowledge gaps and things you're 19 working on. If you can say something about that, I'd 20 appreciate it.
21 Also, it appears you've really looked into 22 the salt usage by the solar plants, and you ought to 23 have pretty good reliability data and the like. I 24 don't see any reason why your tanks and piping systems 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 would be a whole lot different from theirs, except for 1
maybe temperature. Can you talk about that, too.
2 MR. SPRENGEL: So the first question on 3
PIRTs, we've got, I guess I don't know, I don't want 4
to sound too slang usage here, but we do have lots of 5
PIRTs and in several different topic areas. Recently, 6
I'm trying to think, we met with the staff and 7
presented on our use of PIRTs for our core blockage 8
methodology. We've got some other ones in thermal 9
hydraulic uses. So we've got, you know, PIRTs have 10 been fundamental to how we've looked at what, how we 11 model things and then any gaps that we needed 12 additional testing to support.
13 MEMBER MARCH-LEUBA: Are those likely to 14 be part of the docket, or will it be only internal 15 documents that the staff would have to order you see 16 them? Can I see them eventually?
17 MR. SPRENGEL: I don't want to speak for 18 the staff on that. I'm not sure if PIRTs are part of 19 the basics of what we would submit, so certainly they 20 would be available.
21 MEMBER MARCH-LEUBA: See, the fact is the 22 staff, when they need to see one of your internal 23 documents, they just see it. We at ACRS don't have 24 that.
Number
- one, we compress our review 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 dramatically. And number two, we don't have access to 1
those things. So think about the fact of that. A 2
reading room, one of those famous reading rooms, would 3
help our review.
4 MEMBER PETTI: Other applicants have done 5
that. It's very effective.
6 MEMBER BALLINGER: You don't need a PIRT 7
for a construction permit for the salt side, right?
8 I mean, at what point is the boundary, good old 9
fashioned industrial plant where you can build at risk 10 and you don't need a construction permit for that.
11 MR. SPRENGEL: Yes, I don't think we're 12 doing a PIRT for the energy island. It's just the 13 review of operating experience to understand, you 14 know, taking the technology from a concentrated solar 15 plant that had different requirements on reliability 16 to a nuclear plant that wants a high-capacity factor, 17 is there anything that we need to do or is it good 18 enough? That's the investigation, not really a PIRT.
19 CHAIR REMPE: But we do have to do an 20 environmental assessment for the energy island.
21 MEMBER BALLINGER: For anything. What I 22 mean is you don't need a construction permit from the 23 agency to build the salt side, only to a certain 24 point, right?
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83 MR. WILSON: That is correct. This is 1
George Wilson. That is what's in the topical report.
2 It talks about what regulations are applicable or not, 3
so there is a portion of the energy island that would 4
not meet the definition of construction under 50.10, 5
and we would be able to construct it without a limited 6
work authorization.
7 The salt system we're still having the 8
staff look at specifically for portions of it.
9 MR. SPRENGEL: I want to check in.
10 Dennis, I think there was a second part to your 11 question. Did we cover it?
12 MR. BLEY: I think you got it. Oh, yes, 13 I asked have you been able to get pretty good 14 information from the solar folks who use the salt 15 systems, and is there anything really different 16 between your tanks and the things they use?
17 MR. WERNER: Yes, this I Mark Werner.
18 We've been canvassing the concentrated salt power 19 industry quite a bit, and we've got a couple of SNEs 20 on staff that have direct hands-on experience with the 21 systems. So we've been pulling as much knowledge as 22 we can. Being partnered with the DOE through the ARG 23 project has also been helpful because they've got a 24 number efforts through, I think, NREL that we've been 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
84 able to kind of attach ourselves to.
1 And so we are kind of following some of 2
the tank, I'll call them issues that have been 3
cropping up at different plants and feel like we've 4
got a good plan to address the issues that they're 5
currently seeing. I think one thing that we will 6
bring to the table is, you know, we will probably 7
bring a heightened level of quality to the overall 8
project because it is attached to a nuclear plant. I 9
think that will help out for sure.
10 MR. BLEY: I don't have a clue what 11 temperatures they operate at. Are you a whole lot 12 higher than them? What I'm thinking is is information 13 on system performance and reliability going to be a 14 transferrable from the solar plants over to what 15 you're building?
16 MR. WERNER: Yes, it will be directly 17 applicable. Our temperatures are very close to how 18 they operate. I think they might have a slightly 19 higher hot, and we're kind of down at a slightly 20 colder cool, but we overlap very well.
21 MR. BLEY: That's pretty encouraging. So 22 there's nothing really new here.
23 MR. WILLIAMS: Yes. And the salt that 24 we've chosen for this system is this commercial solar 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 salt, so when you hear about solar salt it's a fairly 1
commercial product, and that's what we're using. You 2
know, a lot of the molten salt reactors use a very 3
advanced salt and highly corrosive and whatnot. But 4
this is a well-known salt, and there's lots of it 5
made. It's a commercial product.
6 MR. BLEY: I had a question about the 7
physical nature of the salt. You have really hot 8
liquid salt moving through these pipes. If you get 9
some kind of a break or a leak in a pipe, what's the 10 characteristics of that salt as it comes out? I don't 11 know if salt, if it can flash or what happens to it 12 out there. What's the hazards the operators are going 13 to have to know about and worry about?
14 MR. WILLIAMS: No flashing. It's at 15 atmospheric pressure, so it comes out as a hot liquid 16 and turns into a solid, like that white cake-like 17 substance.
18 MR. BLEY: So pretty benign as it comes 19 out.
20 MR. WILLIAMS: Yes, it is.
21 MR. BLEY: Except if you're touching it.
22 MR. WILLIAMS: Yes, except it's hot. So 23 I should mention, though, our pipes are atmospheric 24 pressure, but they are high temperature. So, you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
86
- know, that definitely involves some design 1
consideration because of that.
2 MEMBER BALLINGER: At 600 C, you're out of 3
the normal stainless steel range. So have you thought 4
about -- the salt people don't use 316 stainless 5
steel. They use other stainless steels.
6 MR. WILLIAMS: Yes, we won't get up that 7
hot.
8 MEMBER BALLINGER: Well, it said 600 C.
9 MR. WILLIAMS: Oh, 600 C, I said stainless 10 steel shows favorable performance with the solar salt 11 up to 600 degrees C. But our core outlet temperature 12 is 510, so nothing is going to be up that hot.
13 MEMBER BALLINGER: I've got 475 stuck in 14 my mind. Certainly, for Section III.
15 MR. WILLIAMS: Yes.
16 MEMBER BALLINGER: Okay, all right. I've 17 got to go back and look and see.
18 MEMBER PETTI: So can you do it in the 19 traditional nuclear part of the code, or do you got to 20 go div 5?
21 MR. WILLIAMS: We go to div 5.
22 MEMBER PETTI: You do go to --
23 MR. WILLIAMS: Yes, we do.
24 MR. SPRENGEL: Looking ahead at our 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
87 planned topical reports, many of them listed here. We 1
continue to work with the staff on these submittals 2
and getting them to review. This month, we're 3
targeting the HFE program plan document, as well as 4
volcanic hazard assessment. And looking ahead into 5
summer, we've got source term, we talked about 6
relation to our EPZ methodology.
7 Later in the fall, our DBA transient 8
methodology, partial flow blockage, and our code 9
usage. And then in the winter, digital I&C and fuel 10 handling I&C. So lots of activity in terms of getting 11 these into the NRC staff's hands for review this year.
12 MEMBER MARCH-LEUBA: I see you have a 13 topical on stability. Are you worried about 14 oscillations in the core, or is it thermal hydraulic 15
-- what in stability are you worried about?
16 MR. WILLIAMS: It's not VNOM transients; 17 I know that. But probably need to call back to the 18 home office.
19 MR.
SPRENGEL:
So the stability 20 methodology is, generally, it's describing how the 21 core functions in steady state, so it's kind of a 22 stable reactor.
23 MEMBER MARCH-LEUBA: Not that it's 24 unstable.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 MR. WILLIAMS: Right.
1 MR. SPRENGEL: Okay. Looking at, we did 2
submit our Natrium engagement plan back in June of 3
2021, coming up on two years of pre-application 4
engagement with the staff, 39 meetings to date. We 5
have seen benefits and have a
good working 6
relationship with the NRC staff. And reiterating here 7
our PSAR content is being developed consistent with 8
the ARCAP guidance, so the structure will be different 9
under the ARCAP guidance from NUREG-0800.
10 VICE CHAIR KIRCHNER: How do you view 11 that, Ryan? Is that just a mapping exercise, or are 12 you the first one going through the wicket, or both?
13 MR. SPRENGEL: Yes, both is probably a 14 better way. The guidance in the form of the groups 15 who have done all the guidance between ARCAP and TCAP 16 and it's been spread out into, you know, who kind of 17 has primary, who has the primary lead on developing 18 that guidance, so it's spread out in many documents, 19 and I think all of it is still draft at this point.
20 So it does lay out the structure of what goes where.
21 I guess I'll give an example of where the 22 first time, you know, evolution has come into play.
23 Some of the structure we've modified and we've worked 24 with the staff and, you know, industry stakeholders in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
89 terms of some modification of where things might fall 1
just to make sense, primarily from a reviewer and 2
packaging standpoint, and have proposed that that's 3
our path forward, and I think some of that is being 4
incorporated, you know, as the draft finalization.
5 We are also going back and we have 6
reviewed regulations and NUREG-0800, the SRP, for any 7
applicability and any kind of gaps that might be 8
there, and we continue to engage with the staff. You 9
know, when we identify something that we see in 10 expectation from NUREG-0800 but we don't see described 11 kind of one way or the other or maybe lightly in the 12 ARCAP guidance, we reach back out and are looking for 13 clarification on the staff of is that intentional that 14 it was actually removed. Most often, that's not the 15 case. And so then we take that on as an additive of 16 17 VICE CHAIR KIRCHNER: That would be my 18 concern because you're doing it as a construction 19
- permit, and that's guided by 50.30 and its 20 requirements. ARCAP is coming along later. That's 21 why I said the first order, the mapping exercise. But 22 then whether the staff -- it's something we'll ask the 23 staff, not, you know, are they ready to map NUREG-0800 24 or some version of NUREG-0800 to the ARCAP guidance, 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
90 so that there's not a lot of -- what I would worry 1
about is not so much that you don't have all the 2
content there, but there's this, like I said, a 3
mapping exercise, and so that the review is efficient 4
for both parties.
5 MR. SPRENGEL: Absolutely. Yes, we are 6
not, we're following the guidance, but we're also 7
doing our due diligence to review the regulations and 8
other sources of kind of, you know, NUREG-0800 is just 9
had more time to develop and be complete and 10 comprehensive, and so we're using the regulation, as 11 well as guidance, to inform and kind of cross-check 12 all those things on our side. I agree it would 13 probably be a fair question for the staff of how 14 they're viewing that, as well.
15 MEMBER BALLINGER: I haven't lost my mind.
16 I'm looking at Table HAA-1130-1 in Division 5, and the 17 upper temperature limit for stainless steel is 425.
18 So I don't know what section we're working to, but 19 Section VIII allows higher temperatures and allows 20 higher allowables. But Division 5,Section III, 21 Division 5, there's a limit there. I saw 600 C on one 22 of your slides, and so maybe I ought to check this.
23 MR. SPRENGEL: Yes, I'll have to check 24 back on the code applicability. I just meant we don't 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 have anything up at that 600 degree C temperature 1
because the core outlet is at 510. We do have it 2
above 425.
3 MEMBER BALLINGER: I mean, Division 5 is 4
640 pages long, and I'm sure you probably missed 5
something.
6 MS. NEIDER: We'll bring that back to our 7
team and evaluate it.
8 VICE CHAIR KIRCHNER: Members, any 9
questions before we take comments from the public?
10 Okay.
11 MEMBER SUNSERI: Just a question, I guess.
12 I heard a lot of the members talk about Part 53, 13 ARCAP, these kind of things. None of that is approved 14 for us, and they won't be licensing to that, so I 15 don't want to send you all a message that we're 16 holding you accountable to Part 53. Are we? We're 17 not, right? But all we've talked about, we've talked 18 several times about that, so I just want to be clear 19 that's not the standard.
20 VICE CHAIR KIRCHNER: That's where I was 21 going, Matt, that, obviously, they have to use Part 22
- 50. That's what they're applying for, construction 23 permit under 50.
24 MEMBER BROWN: The point is don't ask 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
92 questions on the other ones.
1 MEMBER SUNSERI: I mean, it was just 2
sounding to me like it could get confusing to the 3
applicant is all.
4 MEMBER MARCH-LEUBA: I'm going to make my 5
typical talk at the end of this presentation based on 6
this comment that we're asking a lot of questions. We 7
don't have any problem with your design. We like your 8
design, and you're doing a great job, and we like this 9
stage of the project. When we want to start 10 scratching on the surface, then we ask questions that 11 you won't like. But right now, thank you for this.
12 Don't take anything we said as meaning anything 13 detrimental.
14 VICE CHAIR KIRCHNER: And these are 15 comments by individual members at this juncture. This 16 is an informational briefing for us, not a critique of 17 the design.
18 Okay. I'd like to open the floor to any 19 comments from the public. If you're participating 20 online, please mute your microphone, state your name, 21 identify your affiliation if relevant, and make your 22 comment, please.
23 MR. MOORE: This is Scott Moore. You may 24 need to press *6.
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93 VICE CHAIR KIRCHNER: Yes, let me see 1
here. We see a hand is up out there. Oh, Ed, yes, go 2
ahead, Ed. Unmute your mike, please, and make your 3
comment.
4 MR. LYMAN: Yes. I thank you. This is Ed 5
Lyman from the Union of Concerned Scientists. I'd 6
just like to point out that we have some serious 7
safety concerns with fast reactors in general and this 8
design in particular. And in the discussion so far in 9
the open session, you only touched on some of the 10 relevant aspects. And what's frustrating is Natrium 11 does not seem to be transparent. When they talk about 12 the difference between sodium-cooled fast reactors and 13 light water reactors, they rarely mention that the 14 time scale for transients is so much shorter in fast 15 reactors that that is a significant safety flaw. And 16 they emphasize the temperature difference between the 17 operating temperature and the boiling point of sodium, 18 but they don't talk about those transients that could 19 lead to a very rapid increase in coolant temperature.
20 And, yes, this reactor does have a positive void 21 coefficient, and, if it does get to a boiling crisis, 22 you could have a very severe reactivity transient.
23 And the absence of a physical containment 24 is a potential significant safety vulnerability they 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
94 also don't talk about. I know the core assembly 1
events, which I'm sure everyone knows, was a big 2
factor in the Clinch River and FFTF licensing and, as 3
far as I'm concerned, has not been resolved. And so 4
the lack of a strong physical containment is, again, 5
I think a significant open question for this design.
6 So I urge you to explore these questions 7
both in the closed session and also in future open 8
sessions, and that's my comment. Thank you.
9 VICE CHAIR KIRCHNER: Thank you, Ed. Any 10 other members of the public wish to make a comment?
11 Hearing none, then what we will do now is close this 12 Teams' link and go into closed session. Do we need to 13 give any other formalities?
14 CHAIR REMPE: I'd like to suggest that we 15 break until 3:15 to allow the transition to the closed 16 session occur. Also, this is the last that we'll have 17 the open session open today, and so, for those members 18 of the public who want to tune into tomorrow, we'll 19 resume at this link tomorrow at 8:30. Okay.
20 Thank you. And we are recessed.
21 (Whereupon, the above-entitled matter went 22 off the record at 2:54 p.m.)
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
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Plant and Licensing Strategy Overview a TerraPower & GE-Hitachi technology NAT-3292
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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.
2
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Natrium Safety Features
- Pool-type Metal Fuel SFR with Molten Salt Energy Island
- Metallic fuel and sodium have high compatibility
- No sodium-water reaction in steam generator
- Large thermal inertia enables simplified response to abnormal events
- Simplified Response to Abnormal Events
- Reliable reactor shutdown
- Transition to coolant natural circulation
- Indefinite passive emergency decay heat removal
- Low pressure functional containment
- No reliance on Energy Island for safety functions
- No Safety-Related Operator Actions or AC power
- Technology Based on U.S. SFR Experience
- EBR-I, EBR-II, FFTF, TREAT
- SFR inherent safety characteristics demonstrated through testing in EBR-II and FFTF Control Motor-driven control rod runback and scram follow Gravity-driven control rod scram Inherently stable with increased power or temperature Cool In-vessel primary sodium heat transport (limited penetrations)
Intermediate air cooling natural draft flow Reactor air cooling natural draft flow -
always on Contain Low primary and secondary pressure Sodium affinity for radionuclides Multiple radionuclides retention boundaries 3
Control Contain Cool
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Key Differences from Light-Water Reactors 4
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
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Rx Building Fuel Handling Building Rx Aux. Building Control Building Warehouse Warehouse
& Admin Standby Diesels Firewater Steam Generation Turbine Building Switchyard Energy Storage Tanks Salt Piping Single Unit Site Rx Aux. Building Warehouse Warehouse Warehouse
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Reactor Aux. Building Intermediate Sodium Hot Leg Intermediate Sodium Cold Leg Reactor and Core Intermediate Air Cooling Head Access Area Refueling Access Area Reactor Air Cooling / Reactor Cavity Intermediate Reactor Building Fuel Handling Building Reactor Air Cooling Ducts Spent Fuel Pool (water)
Sodium Int. loop Sodium/Salt HXs Salt Piping to/from Thermal Storage System Ground Level 6
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Energy Island Thermal Storage Hot Molten Salt Thermal Storage Tank Cold Molten Salt Thermal Storage Tank Thermal Storage
- Number of tanks based on customers energy need
- Steam generator trains based on size of turbines
- Turbine size based on customers power need Molten Salt Pipe Racks Cold Tank Hot Tank Hot salt pump Superheater Reheater Cold reheat Hot reheat Live steam Steam drum Evaporator Preheater Feedwater Cold salt pump Hot salt from reactor Cold salt to reactor Attemperation pump 7
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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
- Density similar to water
- Lack of corrosion
- Limited auxiliaries
- Sodium inventory - 800 m3 in reactor PCT = Peak Central Temperature CT = Cladding Temperature BP = Boiling Point (atmospheric)
Natrium metallic fuel LWR oxide fuel PCT 1100 C CT 300 C BP 345 C @ 155 atm PCT 650 C CT 550 C BP 883 C @1 atm 8
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Sodium in Liquid and Solid States 9
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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 10
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Salt in Solid and Molten States 11
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Example of Simplified Nuclear Systems 12 Natrium Reactor Air Cooling Zero ASME Sect. III Pipe Welds Atmospheric Pressure (<1 PSI)
Unlimited Air-Cooled Heat Sink Supply Fully Passive (Always in Operation)
Singular Rugged System LWR Emergency Core Cooling 2600+ ASME Sect. III Pipe Welds High Pressure Injection (1000+ PSI)
Large Water Inventory Requirements Active Valve and Pump Operation Multiple Trains and Sub-systems 9799218-13_r0
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Key Features of Reactor Equipment Design
- Pool Type Integral Reactor
- Large Volume of Sodium Coolant
- Atmospheric Pressure
- Separation of Hot and Cold Pool Regions
- Mechanical Pumps
- In-vessel Refueling Reactor Core Heat Exchanger Pump Core Inlet Plenum CRDMs Hot Pool Cold Pool Heat Transport Loop Piping Guard Vessel Reactor Vessel In-Vessel Transfer Machine 13
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Heat Removal 14 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
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15 Submit NRC Application for Construction Permit
( ex PSAR and ER)
NRC holds Public Meeting at the Site Safety Review 10 CFR Parts 20, 50, 73, and 100 Environmental Review 10 CFR Part 51 Construction Permit Issued Submit NRC Application for Operating License (FSAR)
Operating License Issued Fuel Load and Start Up ACRS Review Mandatory Hearings (Comm./ASLB)
ACRS Review Key:
Opportunity for Public Participation Milestone Activity Overview of the 10 CFR Part 50 Licensing Process for the Natrium Advanced Reactor
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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 16
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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 17
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Topical Reports Planned HFE Program Plan and Methodologies Volcanic Hazards Assessment Mechanistic Source Term Methodology DBA Transient Methodology (In-Vessel without Release)
Radiological Release Consequences Methodology Partial Flow Blockage Methodology Engineering Computer Codes for the Natrium Reactor Reactor Stability Methodology DBA Transient Methodology (In-Vessel with Release, Ex-Vessel with and without Release)
Defense-in-Depth and Diversity I&C Strategy Digital I&C (Architecture and Design)
Fuel Handling Instrumentation and Control 18
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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 19
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Questions?
20
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Acronym List ACRS - Advisory Committee on Reactor Safeguards ARCAP - Advanced Reactor Content of Application Project ASLB - Atomic Safety and Licensing Board ASME - American Society of Mechanical Engineers BP - boiling point CFR - Code of Federal Regulations CRDM - control rod drive mechanism CT - cladding temperature DBA - design basis accident EBR - Experimental Breeder Reactor ER - Environmental Report FFTF - Fast Flux Test Facility FSAR - Final Safety Analysis Report HFE - human factors engineering HX - heat exchanger I&C - instrumentation and control IAC - intermediate air cooling system LBE - licensing basis event LMP - Licensing Modernization Project LWR - light-water reactor MWh - megawatt-hour NEI - Nuclear Energy Institute NRC - U.S. Nuclear Regulatory Commission PCT - peak cladding temperature PSAR - Preliminary Safety Analysis Report PSI - pounds per square inch RAC - reactor air cooling system SAR - Safety Analysis Report SFR - sodium-cooled fast reactor SSC - structure, system, and component TREAT - Transient Reactor Test Facility 21
ENCLOSURE 3 Plant Overview Presentation Material - Closed Meeting Non-Proprietary (Public)
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Plant Overview a TerraPower & GE-Hitachi technology NAT-3293
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2 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)
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Reactor Core Overview 4
Core Assembly Reactor Core Reactor Core Map Duct Inlet Nozzle Handling Socket Load pads
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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 5
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Fuel Types 6
((
))(a)(4),ECI
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7
((
))(a)(4),ECI Pin Strip Layer Assembly
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Key Features of Reactor Equipment Design
- Pool Type Integral Reactor
- Large Volume of Sodium Coolant
- Atmospheric Pressure
- Separation of Hot and Cold Pool Regions
- Mechanical Pumps
- In-vessel Refueling 8
Heat Exchanger Pump Core Inlet Plenum CRDMs Hot Pool Cold Pool Heat Transport Loop Piping Reactor Vessel In-Vessel Transfer Machine Reactor Core Guard Vessel
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10
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13 In-Vessel Refueling Refueling Equipment Key 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 14
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Ex-Vessel Fuel Handling Process Overview 15
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16 Ex-Vessel Fuel Handling Process Equipment
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Intermediate Heat Transport Loop 17 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 Commercial Construction Power Cycle Loop (Water)
EI Salt System (Nitrate Salt)
Graded Approach:
Constructed in accordance with NQA Reactor Primary Pool (Sodium) 9799218-7b_r0 hot tank cold tank NI Control Cold pump speed Cold tank level Cold salt temperature EI Salt System (Nitrate Salt)
EI Salt System (Nitrate Salt)
EI Salt System cold cold tank tank EI Control Hot pump speed Hot tank level Hot salt temperature 18
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Reactor Air Cooling 19
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Heat Removal 20 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
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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 21
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Functional Containment - Diverse Barriers 22 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)
Diverse Barriers in Natrium Functional Containment:
Fuel Matrix Sodium Coolant Cover Gas Head Access Area Reactor Building*
Sodium Coolant Reactor Auxiliary Building*
Fuel Handling Building*
Reactor Building*
Environment Environment Environment
- Sub-compartments within building provide additional confinement, depending on event.
Environment Reactor Reactor Environment
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Functional Containment - Analysis and Source Term 23 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
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Site Boundary
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25 Load Following w/ Integrated Energy (Thermal) Storage
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26 Load Following w/ Integrated Energy (Thermal) Storage
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Energy Island Capacity Optimization 28
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Energy Island -
Thermal Storage System 29
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Energy Island - Steam Generator Equipment 30
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Energy Island -
Turbine, Generator and Feedwater Systems 31
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Natrium Hybrid Main Control Room -
Nuclear Island + Energy Island Human Machine Interfaces 32 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 33 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) 35
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Basic SCRAM Design Transient (continued) 36
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Basic Runback Design Transient Sequence of events following a RUNBACK function:
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Basic Runback Design Transient (continued) 38
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39
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Uncontrolled Rod Withdrawal
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Uncontrolled Rod Withdrawal 40
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Uncontrolled Rod Withdrawal 41
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Uncontrolled Rod Withdrawal 42
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Uncontrolled Rod Withdrawal
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- Event sequence (long term response):
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Questions?
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Acronym List 51 AHX - sodium-air heat exchanger ANL - Argonne National Laboratory BLTC - bottom loaded transfer cask CCCS - core component conditioning station CFR - Code of Federal Regulations CRDL - control rod drive line CRDM - control rod drive mechanism DBA - design basis accident DID - defense-in-depth DL - defense line ECI - export controlled information EI - energy island EIC - energy island control system EPZ - emergency planning zone ESS - energy island salt heat transport system EVHM - ex-vessel handling machine EVST - ex-vessel storage tank FTP - fuel transfer port GVDS - group view display system HX - heat exchanger IAC - intermediate air cooling system IHT - intermediate heat transport system ISP - intermediate sodium pump IVTM - in-vessel transfer machine LOCA - loss of coolant accident MCR - main control room NI - nuclear island NIC - nuclear island control system NQA - Nuclear Quality Assurance NSS - nuclear island salt heat transport system P/F - power to flow PIE - post-irradiation examination PRC - pin removal cell PSP - primary sodium pump RAC - reactor air cooling system RPS - reactor protection system RBFP - reactor building floor plug RIS - reactor instrumentation system RV - reactor vessel SFR - sodium-cooled fast reactor SR - safety-related XIS - nuclear instrumentation system