ML24198A111
| ML24198A111 | |
| Person / Time | |
|---|---|
| Issue date: | 06/05/2024 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-2884 | |
| Download: ML24198A111 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
ACRS 716TH MEETING Docket Number:
N/A Location:
Rockville, Maryland Date:
06-05-2024 Work Order No.:
NRC-2884 Pages 1-120 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 716TH MEETING 4
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5
(ACRS) 6
+ + + + +
7 WEDNESDAY 8
JUNE 5, 2024 9
+ + + + +
10 The Advisory Committee met via 11 teleconference at 8:30 a.m., Walter L. Kirchner, 12 Chair, presiding.
13 14 COMMITTEE MEMBERS:
15 WALTER L. KIRCHNER, Chair 16 GREGORY H. HALNON, Vice Chair 17 DAVID A. PETTI, Member-at-Large 18 RONALD G. BALLINGER, Member 19 VICKI M. BIER, Member 20 VESNA B. DIMITRIJEVIC, Member 21 JOSE A. MARCH-LEUBA, Member 22 ROBERT MARTIN, Member 23 THOMAS ROBERTS, Member 24 MATTHEW W. SUNSERI, 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 1
ACRS CONSULTANT:
2 STEPHEN SCHULTZ 3
4 DESIGNATED FEDERAL OFFICIAL:
5 KENT HOWARD 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 P R O C E E D I N G S 1
(8:30 a.m.)
2 CHAIR KIRCHNER: Good morning. The 3
meeting will now come to order. This is the first day 4
of the 716th meeting of the Advisory Committee on 5
Reactor Safeguards.
6 I am Walt Kirchner, Chair of the ACRS.
7 Other members in attendance are Ron Ballinger, Vicki 8
Bier, Vesna Dimitrijevic, Greg Halnon. Expect Jose 9
March-Leuba to join us; Robert Martin, David Petti, 10 Thomas Roberts, and Matt Sunseri. We also have our 11 consultant Steve Schultz on the line virtually.
12 I know we have a quorum today. The 13 committee is meeting in person and virtually.
14 The ACRS was established by the Atomic 15 Energy Act and is governed by the Federal Advisory 16 Committee Act, FACA. The ACRS section of the U.S. NRC 17 public website provides information about the history 18 of this committee and documents such as our charter, 19 by-laws, Federal Register Notices for meetings, letter 20
- reports, transcripts of full and subcommittee 21 meetings, including all slides presented at the 22 meetings.
23 The committee provides its advice on 24 safety matters to the Commission through its publicly-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 available letter reports. Comments by individual 1
members do not represent Committee decisions. The 2
Commission speaks only through its published letter 3
reports.
4 The Federal Register Notice announcing 5
this meeting was published on May 10th, 2024. This 6
announcement provided a meeting agenda, as well as 7
instructions for interested parties to submit written 8
documents or requests for opportunities to address the 9
committee.
10 The Designated Federal Officer for todays 11 meeting is Kent Howard.
12 A communications panel has been opened to 13 allow members of the public to monitor the open 14 portions of the meeting. The ACRS is inviting members 15 of the public to use the MS Teams link to view slides 16 and other discussion materials during these open 17 sessions.
18 The MS Teams link information was placed 19 on the agenda on the ACRS public website.
20 Periodically the meeting will be open to accept 21 comments from members of the public listening to our 22 meeting.
23 Written comments may be forwarded to Mr.
24 Kent Howard, todays Designated Federal Officer 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 A transcript of the presentation portions 1
of the meeting is being kept. And it is requested 2
that speakers identify themselves and speak with 3
sufficient clarity and volume so they can be readily 4
heard.
5 Additionally, participants and members of 6
the public should mute themselves when not speaking, 7
including cell phones, please.
8 During todays meeting the committee will 9
consider the following topics: TerraPower Natrium 10 Topical Reports on Principal Design Criteria, and Fuel 11 and Control Assembly Qualification.
12 And we may get to commission meeting 13 preparations.
14 At this time Id like to ask other members 15 if they have any opening remarks. Members? No?
16 Im not hearing or seeing any.
17 And with that, Im going to turn to Tom 18 Roberts to lead us on in our first topic for todays 19 meeting.
20 Tom.
21 MEMBER ROBERTS:
Thank
- you, Chair 22 Kirchner.
23 Good morning. Today well follow up on 24 two nature and topical reports that were reviewed 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
6 the subcommittee meeting on May 15th.
1 Ill lead a discussion on the topical 2
report for the principal design criteria. And my 3
colleague Dave Petti will lead the discussion on the 4
fuel and control assembly qualification topical 5
report.
6 This was a pretty thorough review in 7
subcommittee. Today we will hear a high level 8
overview and then focus on residual questions from 9
that meeting.
10 For the PDC topical report well focus on 11 technical justification for the approachs plan for 12 functional containment and the application of the 13 SARRDL, or specified acceptable radionuclide release 14 design limit concepts, since both of these seem to be 15 major departures from past practice for Sodium Gas 16 Reactors.
17 For the fuel qualification topical report 18 we reviewed the major pieces of the qualification 19 report and discussed how to support functional 20 containment and the other safety functions of the 21 design.
22 This mornings schedule allows for part of 23 the meeting to be closed to protect TerraPower 24 proprietary and export controlled information pursuant 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 to 5 U.S.C. 552(b)(c)(4). If we need to do this, 1
which I dont expect at this time but well find out 2
depending on the discussion, we will close the public 3
portion of the meeting and then restart the meeting.
4 Ill now turn it over to Candace de 5
Messieres from the NRC staff for any opening comments 6
she might have.
7 MS. DE MESSIERES: Good morning. And 8
thank you for the opportunity to present today.
9 I am Candace de Messieres, chief of 10 Technical Branch 2 in the Division of Advanced 11 Reactors and Non-Power Production and Utilization 12 Facilities in the Office of Nuclear Reactor 13 Regulation, or NRR.
14 Today representatives from TerraPower and 15 the NRC staff will continue discussions from the May 16 15th ACRS Kairos subcommittee meeting on TerraPowers 17 principal design criteria, or PDC, and fuel and 18 control assembly topical reports.
19 Both of these reports are used in 20 reference in the construction permit application for 21 the Natrium Reactor design for Kemmerer Power Station 22 Unit 1 that was recently accepted for detailed 23 technical review by the NRC staff on May 21st.
24 TerraPowers overall licensing approach 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 for the Natrium design follows the Licensing 1
Modernization Project, or LMP, methodology. The 2
Kemmerer Power Station Unit 1 construction permit 3
application represents the first implementation of 4
such an approach in licensing.
5 The PDC topical report describes the 6
result of TerraPowers process to develop PDCs for 7
Natrium using Regulatory Guide 1.232, Guidance for 8
Developing Principal Design Criteria for Non-Light 9
Water Reactors. The topical report was submitted in 10 January 2023, was accepted for detailed technical 11 review in March of 2023, and was the subject of an 12 audit from September to October 2023.
13 The NRC staffs draft safety evaluation 14 was issued on April 12th, 2024.
15 During todays presentation you will hear 16 a summary of key design and regulatory features 17 associated with TerraPowers PDC development approach, 18 including context on the use of a functional 19 containment and specified acceptable system 20 radionuclide release design limits, or SARRDLs.
21 The fuel and control assembly 22 qualification topical report provides TerraPowers 23 plan to qualify fuel and control assemblies for the 24 Natrium Reactor design. The topical report identifies 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 acceptance criteria for fuel qualification and 1
presents select fuel qualification results, in 2
addition to ongoing and planned fuel qualification 3
activities.
4 The topical report was submitted in 5
January 2023, was accepted for detailed technical 6
review in March of 2023, and was the subject of an 7
audit from June through August 2023.
8 The NRC staffs draft safety evaluation 9
was issued on March 20th, 2024.
10 Thank you again for your time and 11 consideration. And we look forward to the discussion 12 today.
13 MEMBER ROBERTS: Thank you.
14 George, are you going to start it?
15 MR. WILSON: We greatly appreciate -- Im 16 George Wilson, Vice President, TerraPower. We greatly 17 appreciate the time of the ACRS to present on our two 18 topical reports for Fuel Qualification and Principal 19 Design Criteria.
20 And with that, Ill turn it over to Ian 21 Gifford.
22 MR. GIFFORD: Thank you very much.
23 My name is Ian Gifford. Im a licensing 24 manager on the Natrium Project. Well start todays 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 discussion with the fuel and control assembly 1
qualification presentation by Dr. James Vollmer. He 2
is participating remotely.
3 James, are you able to hear us?
4 MR. VOLLMER: Yes. I hear you fine.
5 Ready for me to start?
6 MR. GIFFORD: Yes, please.
7 MR. VOLLMER: I think most of you already 8
saw this, so Ill go fairly quickly.
9 MEMBER ROBERTS: Just to clarify, its 10 fine, our intent was to have TerraPower present and 11 then the staff respond. If you care us to go in a 12 different order and present the fuel qualification 13 first, just -- well cover both before we turn it over 14 to staff. Is that right?
15 Okay, thank.
16 MR. VOLLMER: So, Im James Vollmer from 17 TerraPower. Ill provide a quick overview of the fuel 18 and control assembly qualification topical report.
19 Ill go fairly quickly since I think most of you have 20 seen this before. But feel free to slow me down or 21 stop me if Im going too fast.
22 Next slide, please.
23 So, this is a brief high level overview of 24 the Natrium Reactor. Some key features we want to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 call out.
1 That we are using metallic fuel that has 2
been used historically, especially within the DOE 3
program, for Sodium Fast Reactors.
4 Has very high compatibility between the 5
metallic fuel and sodium. Good retention properties 6
of the metallic fuel matrix to retain fission products 7
within the matrix itself.
8 Good compatibility between the two. If 9
there were to be a breach, large thermal inertia for 10 the large sodium pool within the reactor itself to 11 help promote cooling of the reactor with natural 12 convection.
13 And then we also have an additional air 14 cooling passive system for the old reactor vessel to 15 help maintain coolability of the reactor under all 16 conditions, accident scenarios.
17 Likewise, for the control assemblies they 18 are gravity-driven. But then we also have a motor-19 driven control rod runback and scram follow feature as 20 well.
21 And just inherently stable core with 22 increased power or temperature.
23 We will rely heavily on our program from 24 the historic U.S. SFR experience, EBR-I, EBR-II, FFTF.
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 And then we will use, rely on the TREAT tests that 1
were done historically, as well as perform the tests 2
in this reactor.
3 Next slide, please.
4 This is a brief overview on our approach.
5 So, actually, we started our fuel qualification 6
efforts engagement with the NRC in 2019 through a DOE 7
grant, regulatory assistance grant.
8 As part of that, we were relying heavily 9
on NUREG-0800. And interpreted how it applies to 10 Sodium Fast Reactors with metallic fuel. Adapted the 11 requirements specifically to -- well, were directly 12 applicable to metallic fuel, which are not some that 13 were not identified that were, we though, were needed 14 for metallic fuel systems and Sodium Fast Reactors.
15 So, we call these Regulatory Acceptance Criteria.
16 We did submit three White Papers and 17 received feedback from the NRC as part of this 18 process.
19 Next slide, please.
20 So, given the large amount of pre-21 engagement we already had and have our test programs 22 aligned with this, that was the overall structure we 23 used with the topical report since NUREG-2246 actually 24 came out fairly late in our process. But we did 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 include a section directly mapping between our 1
approach and NUREG-2246 to show that we meet all the 2
assessment framework goals identified, except for the 3
one that was not directly addressed was G2.2.1, 4
radionuclide retention requirements.
5 And that was specifically addressed by 6
separate submittal from TerraPowers Radiological 7
Source Term Methodology Report.
8 Next slide, please.
9 So, just a high level overview of the 10 methodology.
11 So, as I mentioned, we identified the 12 regulatory acceptance criteria. For each one of those 13 acceptance criteria we made sure we have a design 14 criteria and the basis for that.
15 And then we included a fuel system 16 description to make sure we can define the fuel system 17 in enough detail that a regulatory can understand the 18 overall design. And thats the basis of our analyses.
19 The design evaluation includes historic 20 operating experience, testing, as well as methods.
21 And then we also included brief sections 22 on testing and inspection of the fuel as well as an 23 ongoing surveillance program within the reactor plant.
24 Next slide, please.
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 A key aspect of it is we did perform PIRT 1
analysis to identify for each individual design 2
criteria. What were the key phenomena that we needed 3
to understand well to ensure that we met the 4
associated limits?
5 So, here are some examples:
6 Thermal creep strain in the cladding is an 7
actual failure criteria that we used. Its purpose is 8
to prevent cladding rupture or coolant flow blockage.
9 And then kind of the key phenomena that 10 influence that criteria:
11 So, the HT9 cladding properties in this 12 particular model. Fuel-cladding chemical interaction, 13 cladding wastage since that thins the cladding wall, 14 and then the fission gas release within the fuel 15 itself because the more the fission gas retained 16 within the fuel matrix and the strain or the stress on 17 the cladding is higher for even more strain.
18 And, likewise, we have the total strain 19 limit that includes the impacts of irradiation and 20 creep swelling on the cladding itself. And, again, 21 thats mainly to preserve coolant channels between the 22 fuel pins.
23 We also have fuel temperature peak fuel 24 cladding or peak cladding temperature, and then 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
15 overall cladding wastage criteria.
1 Next slide, please.
2 So, heres just an overview of our fuel 3
design.
4 So, the center image is the Type 1 fuel 5
pin cross section. So, the green represents the U 10 6
weight percent zirconium that has been tested 7
extensively within the DOE program.
8 The yellow represents the sodium bond 9
between the fuel and the cladding.
Thats 10 intentional, to provide space between the fuel and the 11 cladding so the fuel during irradiation can swell 12 outward and get interconnected porosity that promotes 13 release of the fission gas up to the fuel plenum. So, 14 the sodium is simply there to conduct the heat out 15 until the fuel expands outward to touch the cladding.
16 On the right you see an axial cross 17 section of the fuel pin. So, you have an axial shield 18 slug below the fuel slug. And then the sodium bond 19 actually comes up above the fuel column at the 20 beginning of life. And the fuel expands radially and 21 axially with irradiation. And then the sodium will 22 start backfilling within the fuel once that porosity 23 interconnects.
24 MEMBER PETTI: James.
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 MR. VOLLMER: Yes?
1 MEMBER PETTI: Just a question.
2 That cross section labeled Type 1, thats 3
not to scale, is it?
4 MR. VOLLMER: Not for actual dimensions.
5 MEMBER PETTI: Right.
6 MR. VOLLMER: But to relative scale it is, 7
yes.
8 MEMBER PETTI: Oh. So, like, the yellow 9
is the sodium is as thick as the cladding?
10 MR. VOLLMER: At the beginning of life, 11 yes.
12 MEMBER PETTI: It is. Okay, good. Thank 13 you.
14 MR. VOLLMER: And then heres the 15 hexagonal --
16 CHAIR KIRCHNER: May I follow up, James?
17 This is Walt Kirchner.
18 MR. VOLLMER: Yep. Go ahead.
19 CHAIR KIRCHNER: Yeah. So, just give us 20 a feeling for the performance. When do you expect 21 nominally the fuel expansion to displace the sodium 22 and make contact and then --
23 MR. VOLLMER: Typically -- Go ahead, sir.
24 CHAIR KIRCHNER: And then just explain 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
17 little bit further about how the expansion is 1
accommodated as it goes from radial to axial. Could 2
you just talk through the fuel on this?
3 MR. VOLLMER: Yep.
4 Yeah, so, roughly kind of 2 percent burnup 5
is kind of typical where the fuel contacts the 6
cladding. So, that would be within our first 7
irradiation cycle in the reactor. The fuel would 8
expand radially and make contact.
9 And, again, it also expands axially not 10 exactly at the same ratio but close to the same ratio.
11 And as soon as the fuel makes clad -- contact with the 12 cladding its axial expansion slows down and basically 13 stops at that point. So, limited axial expansion past 14 that initial growth point.
15 And it is largely driven to irradiation 16 growth within the metal, which itself involves fission 17 gas -- excuse me, fission gas pressurizing the fuel.
18 And it really is kind of opposite of Light-Water 19 Reactor fuels where the fuel is very soft, the 20 cladding is very hard so the fuel does behave much 21 more like a putty almost, if you will, to some extent.
22 Does that address your question?
23 CHAIR KIRCHNER: Once you make contact 24 then the, then the further expansion is taking up 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 axially?
1 MR. VOLLMER: It is. But its actually 2
pretty limited because by that point you have a full 3
connect network of porosity within the fuel so that 4
the gas is released to the plenum at that point in 5
time. So, your driving force to expand is reduced.
6 So, it doesnt keep growing axially typically beyond 7
that point.
8 CHAIR KIRCHNER: Thank you.
9 MR. VOLLMER: Yep.
10 Next slide.
11 So, these images, again, are attempted to 12 be to scale to each other. So, it kind of shows the 13 EBR-II cross section of the fuel pin. The larger 14 metallic fuel, the MFF fuel assemblies, and FFTF, and 15 then the Natrium Type 1 fuel.
16 And, so you do see the Type 1 is slightly 17 larger than the MFF fuel but it is within what has 18 been tested historically in other metallic fuel test 19 designs.
20 You also see heres a cross section on the 21 right side of height between the different fuel 22 assemblies. You do see that the FFTF fuel column is 23 much taller than the EBR-II. And although the Natrium 24 Type 1 overall fueling fuel column, fuel -- 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
19 fuel assembly link is much taller, the fuel height is 1
actually almost exactly the same as the FFTF fuel 2
column. And that was intentional that we did not want 3
to extrapolate beyond what was tested for the overall 4
fuel height.
5 Next slide, please.
6 So, from a fuel system design evaluation, 7
we are relying heavily on the historic operating 8
experience because there are no longer any Fast-9 Operating Reactors in the U.S., plus there is a wealth 10 of historic data that we were able to rely on. So, we 11 have been working for many years with the DOE labs to 12 obtain legacy fuel data as well as qualify it.
13 Not only does this include full fuel pin 14 irradiation tests but also has fuel, and material 15 properties and transient and accident tests that we 16 were able to use.
17 We do have several other test activities 18 in progress or planned, including, so the FFTF fuels 19 are most reflective of our fuel design. Most of those 20 were not looked at after irradiation. So, we have 21 sponsored additional post-irradiation exams to address 22 some of the gaps in the historic database and just 23 demonstrate its performance relative to the EBR-II 24 fuels to get a fuller picture and understanding.
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20 Also, transient testing of fuel pins. So, 1
the TREAT reactor in Idaho was specifically designed 2
for severe accident testing of fuels to test to 3
failure. That has been re-started recently, so we do 4
plan on testing full length irradiated metallic fuel 5
pins from the FFTF reactor, and then also additional 6
furnace tests to just better characterize the 7
transient behavior in severe accidents.
8 We have fuel and absorber property tests, 9
including we created metallic SIMFUEL, so we simulate 10 burnup in the fuels by adding representative fission 11 product species to it.
12 We have a host of HT9 materials tests.
13 And then core assembly and mechanical 14 tests as well.
15 Next slide, please.
16 So, for our materials test programs we 17 actually kicked these off 2011 time frame. Our first 18 step was actually to get HT9 materials. So, we 19 actually worked with multiple suppliers and got three 20 unique heats of HT9 just to characterize the bounds of 21 the specification, as well as understanding the 22 impacts to performance.
23 We actually chose those compositions based 24 on the historic operating experience of the DOE 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 materials and looking at those materials.
1 Some of the key gaps we felt from the 2
historic database was expanding the time at 3
temperature for those. We did an extensive thermal 4
aging program up to 50,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> just to understand 5
are there any microstructural changes in it just due 6
to time at temperature?
7 And then doing microstructural 8
characterization and mechanical testing on those.
9 Again, we did actually make new heats of 10 HT9 material just because we wanted to verify how it 11 performed relative to historic HT9s. We have side-by-12 side irradiations of that material with some of the 13 legacy DOE material in the BOR-60 reactor in Russia, 14 and have it irradiated up to about 85 dpa for that.
15 We also have planned irradiation tests on 16 welds and coatings, and some advanced materials. This 17 is on the High FIR Reactor in Oak Ridge National Lab.
18 And then thermal creep testing is another 19 kind of long time at temperature phenomena we were 20 concerned about. So, we have tested up to 70,000 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> for thermal creep testing for HT9. And then 22 also have axial tube creep underway and biaxial tube.
23 And really the purpose of these tests is 24 to help us refine our overall response models for 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
22 HT9 behavior.
1 MEMBER PETTI: James, in terms of the 2
radiations in BOR-60.
3 MR. VOLLMER: Yep.
4 MEMBER PETTI: Theyre complete? Are the 5
samples back in the States yet?
6 MR. VOLLMER: Not yet. We are close to 7
shipping them back. I think theyre in the process of 8
packing and queuing them up right now.
9 MEMBER PETTI: Okay. Thats good.
10 I remember when it started in the DOE 11 program.
12 MR. VOLLMER: Yeah. Yeah, its been a bit 13 of a journey to get them back.
14 MEMBER PETTI: A long time coming.
15 MR. VOLLMER: Yes, yes. But were close.
16 MEMBER PETTI: Good.
17 MR. VOLLMER: Next slide.
18 So, this is just a brief overview of some 19 of our fuel performance tools at TerraPower. So, from 20 a fuel pin performance point of view.
21 So, we have two codes crucible. Its a 22 fast-running code thats actually integrated with our 23 overall core design software, the ARMI software. And 24 it really is aimed at the key phenomena that 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
23 tightly coupled with neutronic responses for, like, 1
the fission gas
- release, the sodium-bond, 2
infiltration, fuel axial growth.
3 But we also do have some of the fuel 4
performance phenomena like cladding wasting --
5 cladding wastage, clad temperatures, cladding strain.
6 So, as they are iterating the core, they can verify 7
the ARMI where we expect to meet our fuels design 8
criteria as part of that process.
9 But once they have the overall core 10 designs that they think meets all those goals, then 11 they will give us individual assembly fuel pin 12 histories so that we can perform our detail analysis 13 with our ALCHEMY Package, which is a finite element 14 base method.
15 On the right is an example of a cladding 16 tube where you can see the different finite elements 17 and the actual predicted strains along that fuel pin.
18 Again, it is a high-fidelity model.
19 Captures all the phenomena we think are key for 20 modeling metallic fuel behavior, fission gas release, 21 FCCI, thermal conductivity. But then, in addition to 22 fuel, we are able to adapt models for the boron 23 carbide as well, so we can use the same tool for our 24 absorber predictions as well as our fuel predictions.
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24 And then we also have models specifically 1
designed to support our ongoing irradiation tests so 2
we can pre-test predictions ahead of time, run the 3
actual tests, and then analyze the results and 4
compare. So, it helps validate the model in the 5
process as well.
6 Also, the ALCHEMY Package is used as kind 7
of the structural material models for the ATR material 8
that are used at our higher linked scales, in 9
particular OXBOW for our full fuel assembly models and 10 for restraint system. And the materials come from 11 ALCHEMY for those.
12 Next slide, please.
13 So, I mentioned OXBOW. Thats our primary 14 core mechanical performance tool. Can do single 15 assembly just to verify kind of the amount of 16 distortion anticipated within a fuel for core 17 assembly. And then prediction, kind of withdrawal and 18 insertion tech loads based on those distortions. But 19 then also from a core-wide, core lockup response as a 20 function of thermal or irradiation behavior, can use 21 the same tool.
22 And then also perform it for seismic 23 analysis as well.
24 We also have a module within OXBOW 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
25 also do control assembly scram under seismic tech 1
response and assembly drop times, control assembly 2
drop time. And then bundle-duct interactions as well.
3 Next slide, please.
4 We have extensive testing underway for our 5
core assembly response. On the right side this is 6
actually a fixture we use for the mechanical testing.
7 We can actually fit multiple fuel assemblies within 8
that.
9 In the center picture, thats actually 10 within our Bellevue Lab of a pit with that inside of 11 it where full length fuel assemblies can be distorted, 12 put in there, measure the withdrawal insertion loads 13 to pull them in and out.
14 Likewise, we can load multiple assemblies, 15 apply thermal gradients, verify the bending response 16 of them.
17 On the bottom right shows a sample of kind 18 of a bundle compression test just to look at how does 19 the bundle redistribute with loads applied from given 20 bases of the assembly.
21 Have a whole host of kind of single 22 assembly, multiple assembly, and then these bundle-23 duct type interaction tests as well.
24 And then we also have worked with 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
26 international community to benchmark against historic 1
codes as well as historic databases for operating Fast 2
Reactors.
3 MEMBER PETTI: Then, James, just a 4
question on that.
5 MR. VOLLMER: Yes.
6 MEMBER PETTI: The rest of the world is in 7
oxide space.
8 MR. VOLLMER: Yes.
9 MEMBER PETTI: Is it still valuable? I 10 mean, youve got then the metal system vs. the oxide 11 system?
12 MR. VOLLMER: Yeah, very much so, that the 13 thermal gradients within the fuel assemblies are 14 largely the same and have the exact same radiation 15 effects. And I did see DOE HT9 material, so it was 16 actually an oxide fuel assembly achieved the highest 17 DPA on. So, weve been using that assembly 18 extensively for benchmarking for the dilation and 19 whatnot due to radiation performance.
20 So, its very relevant overall.
21 MEMBER PETTI: Okay. Thanks.
22 MR. VOLLMER: Yeah.
23 Next slide, please.
24 Kind of really the last piece of 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
27 qualification program is our fuel surveillance 1
program.
2 So, we did actually design Type 1 fuel to 3
be very conservative relative to its historic designs.
4 And believe there is enough margin in lifetime we can 5
connect it to an additional cycle. But weve 6
restrained it to what the historic operating 7
experience was just to verify we are bound by history.
8 But we do have special fuel assemblies, we 9
call them our Lead Demonstration Assemblies, that do 10 have pins that we can remove in these, the X fuel 11 handling to pull them out for expedite post-12 irradiation exams. So, that way we can constantly 13 monitor performance throughout lives.
14 So, after each cycle be able to pull pins 15 out, do visual exams on them, measure them, send them 16 off for extensive post-irradiation exams or structural 17 exams just to make sure that the fuel is behaving as 18 predicted based on the historic operating experience.
19 We will be targeting a subset of them to 20 actually have accelerated burnup so well actually 21 maximize the enrichment of those pins so that they are 22 dating the rest of the core as far as burnup.
23 And then also trying to target bounding 24 conditions for some of them as well just to verify 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 that we are bounding the entire performance of the 1
reactor.
2 Next slide, please.
3 Yeah, I think thats it for me. So, any 4
questions?
5 MEMBER ROBERTS: Yes, a question.
6 If you can go back to slide 6. This slide 7
is a, its how the PIRT Evaluation Identifies Fuel 8
Phenomena.
9 MR. VOLLMER: Yes.
10 MEMBER ROBERTS: And it appears to be a 11 list of design limits --
12 MR. VOLLMER: Yes.
13 MEMBER ROBERTS: -- on the fuel.
14 MR. VOLLMER: Yes.
15 MEMBER ROBERTS: Such that when you do the 16 analysis, either safety analysis or steady state, you 17 would go verify those five limits are met. Is that 18 right?
19 MR. VOLLMER: Correct. Correct.
20 MEMBER ROBERTS: So, that sounds like you 21 would transition to the PD2. It sounds like a SAFDL.
22 MR. VOLLMER: Thats correct.
23 MEMBER ROBERTS: Just wanted to understand 24 what the difference was. If you intended to meet all 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
29 these five limits and then call that a SARRDL, Im 1
just trying to understand what the, what the overall 2
intent is.
3 Because it sounds like your intent is for 4
the design to meet these five limits.
5 MR. VOLLMER: Thats we do design to them 6
and then what happens then goes off to SARRDL space.
7 So, thats where we have our damage criteria. That 8
would be basically to say thats when the fuels 9
reached its effective lifetime. So, if you go through 10 an AOO, and then beyond that we can see the fuel 11 damage, you would not reuse the fuel past that point.
12 But then we do have failure criteria. So, 13 if you did exceed that, we would say the fuel has been 14 failed. And then that would go over to SARRDL space 15 for them to propagate whats the impact of that 16 failure.
17 MEMBER ROBERTS: So, for normal operation 18 AOOs, which is what the JDC or PDC states, you would 19 expect to have a zero release by having met these five 20 criteria? Is that what would be here?
21 MR. VOLLMER: Exactly.
22 MEMBER ROBERTS: Okay.
23 MR. VOLLMER: Exactly, yep.
24 MEMBER ROBERTS: And then for more severe 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 events you would simply calculate the amount of 1
damage. Thats not a SARRDL, right, thats a design 2
basis calculation of what the consequence is; is that 3
right?
4 MR. VOLLMER: So, I think both. We 5
basically calculate how many fuel pins are potentially 6
failed. And then it would go on the SARRDL space to 7
understand what would be the dose potentially released 8
from that.
9 MEMBER ROBERTS: Right. Release in a 10 containment and then the whole, you know, down in 11 those type conditions.
12 Okay, thanks.
13 MR. VOLLMER: Any others?
14 MEMBER PETTI: So, just to make sure Im 15 clear. Then you really have both SAFDLs and SARRDLs, 16 depending on the space of the, of the accident domain, 17 if you will.
18 MR. WILSON: This is George Wilson from 19 TerraPower.
20 Well discuss this more in the PDCs. So, 21 if youll wait till we get to the PDCs --
22 MEMBER PETTI: Perfect.
23 MR. WILSON: -- and ask additional 24 questions well go into a little more detail.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
31 MEMBER PETTI: I have another. I have a 1
question on the surveillance stuff, which I think was 2
really good.
3 How often do you anticipate pulling the 4
pin out of that test assembly?
5 MR. VOLLMER: For the initial cycle --
6 MEMBER PETTI: Right.
7 MR. VOLLMER: -- through the entire fuel 8
lifetime, every cycle well be pulling a subset of 9
pins out. And I guess just to clarify, were not 10 putting the fuel assembly back in after we pull it 11 out.
12 MEMBER PETTI: Back in, right. I figured 13 that, not replacing it.
14 MR. VOLLMER: Yes.
15 MEMBER PETTI: But do you have to put 16 something back in the core, though? Just, or there 17 would just be a hole?
18 MR. VOLLMER: Well, just replace it with 19 a fresh fuel assembly.
20 MEMBER PETTI: Oh, you put it back.
21 Right. Got you.
22 Okay. Thanks.
23 MR. VOLLMER: Yep.
24 CHAIR KIRCHNER: Jim, just on that topic.
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 Will those be wire wrapped or theyll be kind of a 1
straight pin? How do you compensate for the non-2 prototypicality that you get out of those?
3 Is there any issues that you are seeing of 4
not, not having that pin that you removed wire 5
wrapped?
6 MR. VOLLMER: They will still have the 7
neighbors will have wire wraps. So, they will still 8
have the support. It will be a small perturbation.
9 And we are doing thermal hydraulics testing just to 10 verify what it is. But we do anticipate it would be 11 a fairly small difference on it.
12 If anything, theyll likely have a little 13 more propensity to move. So, a little more spreading 14 type interaction.
15 CHAIR KIRCHNER: Right.
16 MR. VOLLMER: But, again, we expect it to 17 be very small, just that the bundle is so tight that 18 there really is not much room for movement.
19 Yeah, were doing harmonic testing to 20 verify that.
21 CHAIR KIRCHNER: Okay, thank you.
22 MR. VOLLMER: Yep.
23 MEMBER PETTI: Just another question comes 24 to mind.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 What irradiation testing of the welds, --
1 MR. VOLLMER: Yes.
2 MEMBER PETTI: -- coatings, and like, 3
youre not going to get the DPA. Is there some 4
historic data so that you can kind of make 5
correlations based on, you know, this stuff looks as 6
good as the old stuff, so we can use the old, the old 7
stuff?
8 MR. VOLLMER: Yeah. We do. Theres a lot 9
of kind of TIG welding I think historically was 10 primarily used. We are wanting to use a different 11 welding process. But it should have a smaller heat-12 affected zone and whatnot.
13 But also, the welds typically are out of 14 the high flux area.
15 MEMBER PETTI: True. Yeah, right.
16 MR. VOLLMER: They dont receive much of 17 the dose as well.
18 MEMBER PETTI: What sort of dose are you 19 going to get?
20 MR. VOLLMER: For the welds?
21 MEMBER PETTI: Yes.
22 MR. VOLLMER: For most of them it will be, 23 I think, less than 5 gpa, as I recall.
24 MEMBER PETTI: Thats my guess. That is 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 what I guessed. Okay.
1 MR. VOLLMER: The control assemblies, that 2
is the one where it will actually be in the higher 3
flux region of the core. So, that was kind of the 4
area we want to make sure we do a bound aspect of it.
5 CHAIR KIRCHNER: Jim, what do you think --
6 this is Walt again -- what do you think of your list 7
of parameters youre testing for, what do you feel is 8
the kind of the limit for your, you know, your fuel 9
design for this, for this application?
10 MR. VOLLMER: So, FCCI is kind of the 11 most, like, limiting in our experience, continues kind 12 of the hot channel factors. So, thats why we spent 13 the most effort for our post-irradiation exams we are 14 doing on the FFTF pin, but specifically looking at the 15 FCCI response which we think that is, again, likely to 16 limit.
17 Because the DOE fuels were high enriched, 18 so they would be higher linear power since theyre 19 lifetime faster. So, kind of the time at temperature 20 combination. So, thats really what we want to make 21 sure we understand that phenomenon well.
22 CHAIR KIRCHNER: Thank you.
23 MR. VOLLMER: Yep.
24 MEMBER PETTI: Just one more since youve 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 got such a wealth of knowledge.
1 All of the out of trial thermist tests 2
that I remember reading the report that gives you the 3
rate of attack, if you will, in that plot I can 4
remember, does the in-pile stuff agree generally well 5
with the out-of-pile there, I mean given all the 6
uncertainties?
7 MR. VOLLMER: For the FCCI, the historic 8
models found no effect of irradiation. Weve actually 9
for our SIMFUEL stuff we found the out-of-pile to be 10 more aggressive than the in-pile, but we believe 11 thats just because we havent been able to recreate 12 the mixture of the fuel products quite like was the 13 actual irradiated fuel itself.
14 MEMBER PETTI: So, its going to serve it?
15 MR. VOLLMER: Right.
16 MEMBER PETTI: I mean, there are examples 17 in LWR space where they tried to make fuel, simulated 18 fuel, and it was found to be grossly over-conservative 19 than --
20 MR. VOLLMER: Yes.
21 MEMBER PETTI: -- what you see in-pile.
22 But, you know, that took two years to get there. And 23 look back and that was a really dumb idea, you know.
24 You understand why we do it, because its easier. But 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 you run that risk.
1 Okay, good.
2 MR. VOLLMER: Yeah, were relying strictly 3
on the in-pile data for our FCCI correlation. The 4
out-of-pile is more to try to understand 5
mechanistically if we can refine the model, or just to 6
give us more insight into how to model the data.
7 So, yeah, we kind of gave up on using it 8
from a purely mechanistic under pure correlation.
9 Just want to do more just a qualitative insight of the 10 behavior from our out-of-pile tests.
11 Its time for me to turn it over to Ian.
12 MR. GIFFORD: Thank you, Jim.
13 So, Ill provide a brief overview of the 14 methodology that was used to develop the principal 15 design criteria for the Natrium Advanced Reactor. And 16 then Ill turn it over to Eric Williams, who is our 17 senior vice president and design authority, for a 18 focused discussion on SARRDL and functional 19 containment.
20 The approach to PDC development was 21 discussed with NRC staff during public meetings in 22 December of 2021, and November of 2022. And the PDC 23 topical report was submitted in January of 2023.
24 In accordance with 10 C.F.R. 50.34, 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 Principal Design Criteria were also included in the 1
construction permit application, Section 5.3.
2 Regulatory Guide 1.332 provides guidance 3
for Non-Light Water Reactors to develop principal 4
design criteria for Non-Light Water Reactor design.
5 The Reg Guide acknowledges that different requirements 6
may need to be adapted for Non-Light Water Reactor 7
designs, and that the PDC in 10 C.F.R. Part 50, 8
Appendix A, are you regulatory requirements for Non-9 Light Water Reactor designs, but they provide 10 guidance in establishing the PDC for Non-Light Water 11 Reactor designs.
12 Ultimately, its the responsibility of the 13 applicant to development PDC for its facility based on 14 the specifics of its unique design.
15 Applicants are allowed to use the Reg 16 Guide to develop all or part of the principal design 17 criteria, and are free to choose amongst the Advanced 18 Reactor design criteria, Sodium-Cooled Fast Reactor 19 design criteria, or Modular High Temperature Gas 20 Reactor design criteria to develop each piece.
21 PDC were developed starting with the 22 SFRDC. And in Appendix D of Reg Guide 1.232, first 23 discussion was whether the PDC applied. And if it did 24 it was assessed for whether it could be adopted as 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 written.
1 If it could be adopted as written, it was 2
accepted as an initial Natrium PDC.
3 If it was needed to be modified, we first 4
reviewed the ARDCs and the MHARDCs for language that 5
may be more applicable to our design. And we also 6
left open the option that we may in fact have to draft 7
new PDCs for Natrium.
8 I want to focus a little bit on the box 9
here. So, we have the initial Natrium PDC list and 10 then the box that says perform the iterative LMP 11 process.
12 So, the LMP is an iterative process 13 throughout the design phase. NEI 18-04, states that 14 Reg Guide 1.232 should be used as an input by 15 designers to initially establish principal design 16 criteria for the facility based on the specifics of 17 the design 18 And then, as part of the LMP process, PRA 19 safety functions are identified that are necessary and 20 sufficient to meet the frequency consonance target for 21 all design basis events and high consequence beyond 22 design basis events to conservatively ensure that 10 23 C.F.R. 50.34 dose requirements can be met.
24 The PRA safety func -- these PRA safety 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 functions are then defined as required safety 1
functions, or RSFs. RSFs are used to develop required 2
functional design criteria, RFDCs, that establish 3
reactor-specific functional criteria that are 4
necessary and sufficient to meet the required safety 5
functions.
6 NEI 18-04 states that the required 7
functional design criteria, RFDCs, are defined to 8
capture design-specific criteria that may be used to 9
supplement or modify the applicable GDCs or ARDC in 10 the formulation of principal design criteria.
11 The Natrium Project has undergone a 12 complete iteration of LMP to include a thorough review 13 by the Integrated Decision Making Process Panel. In 14 accordance with NEI 18-04, the Natrium LMP Design 15 Criteria Report includes a complete mapping of LMP 16 evaluated functions.
17 RFDCs developed from the LMP are all 18 mapped to at least one principal design criteria, 19 demonstrates that the PDCs are complete with the 20 current Natrium design. No RFDCs were found that 21 would require a new or expanded principal design 22 criteria.
23 As we progress the design and continue 24 with LMP, the completeness of the PDC will continually 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
40 be revisited. Significant changes to the principal 1
design criteria are not expected, but should they be 2
needed it would be appropriately communicated to the 3
NRC.
4 MEMBER MARCH-LEUBA: This is Jose.
5 Let me emphasize what you just said Im 6
100 percent Im in agreement with. But when during 7
that slide it says we are starting with a set of 8
design concepts, and then we remove the ones that 9
dont apply.
10 We need to see is there something special 11 with my system that requires a new one. The thing 12 that comes to mind is now theyre available to us 13 through the source.
14 So, Id like it that youre seeking that 15 one.
16 MR. GIFFORD: Appreciate the comment.
17 Thank you.
18 MEMBER PETTI: So, then its fair to say 19 that the PDC Report that were reviewing represents 20 the final PDCs in the bottom box or the initial PDCs?
21 Or are they the same based on where you are?
22 MR. GIFFORD: They are the same based on 23 where we are.
24 MEMBER PETTI: Great.
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41 MEMBER ROBERTS: I have a similar 1
question, is the topical report doesnt have those 2
bottom two boxes. And there is a limitation in 3
condition in the safety evaluation that does tie to 4
that a little bit. It says that the NRC acceptance is 5
based on an understanding you are using LMP process, 6
so that pretty well drives what you just said.
7 Is there an intent to change the topical 8
report or is the reference from the SCR considered to 9
be sufficient to ensure that a future user does what 10 you just said and not what the topical report says?
11 Let me pull this up. It has this figure 12 without those bottom two boxes. And theres no 13 discussion, at least that I recall, in the report 14 itself of that LMP iteration.
15 I believe I think its very important.
16 And it sounds that you do, too. So, its just a 17 matter of making sure whoever uses the topical report 18 understands what you just said.
19 MR. GIFFORD: Yes. The intention would be 20 whoever uses this topical report would be using NEI 21 18-04. And so, following 18-04 would require that an 22 applicant would go through those iterative steps.
23 I think the intention of the figure in the 24 topical report is to show how the PDC were developed 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
42 at the time that the topical report was submitted in 1
January 2023. And then we have subsequently followed 2
the LMP process and moved that into the construction 3
permit application.
4 MEMBER ROBERTS: Okay, thanks.
5 MR. GIFFORD: At this time I will turn it 6
over to Eric Williams for a discussion on SARRDLs and 7
functional containment.
8 MR. WILLIAMS: All right, thank you.
9 So, my name is Eric Williams, Senior Vice 10 President and Design Authority, TerraPower.
11 I was, you know, reflecting on the 12 questions that were asked in the last meeting and 13 trying to, you know, come up with the best approach to 14 try and get some clarity behind these issues. And, 15 you know, coming at it from the design perspective I 16 first just wanted to mention a couple of the new 17 things that TerraPower is doing in the design of 18 Natrium I think are incorporating a lot of this new 19 material on SARRDLs and functional containment.
20 So, you know, using the LMP approach is 21 the first thing, you know, thats sort of new here, 22 following the risk-informed performance-based 23 approach. And then the use of something like a SARRDL 24 becomes really integrated closely with that approach.
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43 And so, talking about them in separate conversations 1
becomes really hard.
2 Were also following a systems engineering 3
approach to design. So that means that were trying 4
to rigorously set functional design requirements at 5
the beginning, including safety requirements that all 6
of these things factor into so that the designers can 7
have clear requirements as they go through their 8
iterations and know that theyre meeting an acceptable 9
design.
10 And then, finally, were using the IAEA 11 framework that calls out defense line functions where 12 it allows us to look at design requirements in each of 13 those defense lines, so that as the designer is moving 14 through their work they can also be evaluating 15 defense-in-depth adequacy as well.
16 And
- then, of
- course, through the 17 Integrated Decision Making Process Panel thats part 18 of LMP, we get a chance as a group to review that 19 defense-in-depth and how were meeting the frequency 20 consequence limits with margin in the design.
21 So, what you get out of that is a really 22 great integrated set of design and safety 23 requirements. But it also means that these things are 24 hard to pull apart from each other.
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 So, thats kind of the framework that 1
were applying. And so, Ill start by talking about 2
SARRDLs first and then that will roll into functional 3
containment. And theyre really closely connected.
4 So, weve already talked a little bit 5
about SARRDLs.
6 So, what is a SARRDL? First thing we do 7
is we look at all the systems and components that are 8
containing radionuclide inventory in the plant.
9 Sometimes storing it, sometimes circulating it during 10 operation. And were trying to set clear limits on 11 potential releases for that circulating radionuclide 12 inventory during normal operations or AOOs.
13 So, we look at each one of those 14 radionuclide-containing systems. And if it has the 15 potential to violate a release limit, then it gets a 16 SARRDL. And that SARRDL is usually in the form of a 17 volumetric leakage rate that can happen from that 18 system.
19 And so what it allows us to do is set a 20 clear design requirement on those SSCs that get the 21 SARRDLs so the designers can use those in design and 22 know that theyre meeting the requirements.
23 Its also very convenient to establish 24 design -- or analysis assumptions. So, when youre 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 looking for the worst case initial condition to 1
initiate, say, a DBA analysis, you can initiate it 2
assuming that youre at the SARRDL limit in that 3
system.
4 So, thats a convenient way. And it makes 5
me think back to Light Water Reactors that may assume 6
youre at your worst case tech spec limit of primary 7
system activity when you initiate an accident. So, 8
thats a really convenient way to do it.
9 And then, in the end those SARRDLs get 10 incorporated into the frequency consequence curves 11 that you see with LMP that ultimately show with 12 specific PRA datapoints showing up on the F-C curve 13 with uncertainties identified that shows that you have 14 margin to the F-C curve limits.
15 And that ultimately is a practical way to 16 demonstrate the PDC 10 compliance where the SARRDLs 17 are mentioned.
18 So, thats how the SARRDLs are used. And, 19 you know, the SECY paper 18-0096 talks about how those 20 are closely linked with functional containment because 21 they express the limits at a performance criteria for 22 functional containment for the AOOs.
23 So, we can talk a little bit more about 24 functional containment now. And I wanted to kind 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
46 start with a couple of the design features because I 1
know one of the questions was really talking about the 2
rationale --
3 MEMBER ROBERTS: Excuse me, Eric.
4 MR. WILLIAMS: Oh, sure.
5 MEMBER ROBERTS:
GDC 10, just to 6
understand how that works with the SARRDL concept.
7 What we heard in the previous presentation 8
is the fuel limits are going to be tracked as design 9
limits. So, essentially you have no release from fuel 10 during normal operation or AOOs. But then you said 11 that the SARRDLs become other circulating activity and 12 other radiation-containing systems, --
13 MR. WILLIAMS: Yes.
14 MEMBER ROBERTS: -- not the fuel.
15 MR. WILLIAMS: Uh-huh.
16 MEMBER ROBERTS: That seems like a 17 different interpretation of that PDC -- or GDC. The 18 purpose of that GDC seems to be that you not have any 19 challenge to radionuclide release from AOOs or normal 20 operation by keeping all the circulating activity 21 within the fuel.
22 Sounds like youre doing exactly that.
23 But then youve added, because of, I guess, the LMP 24 and the need to have a downwind dose reduced in normal 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 operations and AOOs, expanding what GDC 10 is, PDC 10 1
currently says to include circulating activity. Is 2
that the right interpretation?
3 MR. WILLIAMS: Let me try and think about 4
that.
5 MEMBER ROBERTS: Because it seems like the 6
intent of that PDC is to not have release from fuel 7
during normal operation or AOOs.
8 MR. WILLIAMS: Its really to not have any 9
releases that violate the 10 C.F.R. 20 limits that are 10 imposed on normal operation and AOOs by establishing 11 clear requirements for all the radionuclide-containing 12 systems.
13 If we meet the fuel design limits 14 perfectly, then there wont be any from the primary 15 system.
16 There can also be, you know, radionuclide 17 inventories within systems like sodium processing 18 systems, sodium cover gas from some prior failed fuel 19 that occurred, you know, even just randomly. So, you 20 have to look at those.
21 You have refueling systems that have to be 22 looked at, too, because they contain fuel at times.
23 And so, the SARRDLs look at all of those 24 systems and incorporate these limits, not just in 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
48 primary system.
1 But we do use the design limits, as James 2
was pointing out. We look at those in terms of the 3
mechanistic source term methodology. So, it would 4
determine the failed fuel fractions that would get 5
incorporated into mechanistic source term.
6 But the real intent of the SARRDL is to 7
capture what is really the phenomena that is really 8
important for a Sodium Fast Reactor. Since theres 9
not a direct, as direct coupling from fuel fraction, 10 fuel failure fraction to radionuclide release, because 11 we have all these extra systems that act to, you know, 12 attenuate radionuclide release, we have to incorporate 13 all of that. And the SARRDL does that.
14 And so, its really a better metric of 15 what is happening in a Sodium Fast Reactor. And 16 thats kind of the intent of this is to address it 17 directly on whats happening.
18 But were still using the fuel design 19 limits, like you said, as part of the mechanistic 20 source term process.
21 MEMBER MARTIN: Eric, this is Member 22 Martin.
23 You know, what you describe doesnt sound 24 any really different than what weve always done.
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49 Its always been a kind of analytical defense-in-1 depth. You know, the ultimate metric is the SARRDL; 2
right? Were interested in doses and its comparison 3
to 10 C.F.R. 100.
4 Thats backstop.
In fact we can 5
demonstrate the SAFDLs. We have high confidence on 6
dose. And then those SAFDLs are backed up by tests.
7 Weve always done it that way.
8 I guess Im not seeing something new here, 9
except for maybe a documentation emphasis in SARRDLs.
10 And maybe, and thats consistent with LMP, no doubt.
11 But in practice, which I think makes LMP practical, is 12 that its not a big departure from what we do.
13 Obviously, LMP brings in a lot of the risk aspects.
14 And so, its another way of defending any 15 kind of engineering judgment, you know.
16 MR. WILLIAMS: Yeah.
17 MEMBER MARTIN: But Im not really seeing 18 anything new here.
19 MR. WILLIAMS: Yeah.
20 I dont think of it as that new, other 21 its more integrated because in the LMP approach 22 youre going to see all of the results together in 23 this frequency consequence curve. And the SARRDLs are 24 consistent with the way DBAs are also looked at, 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
50 DBEs are looked at.
1 So, it all forms this, like, self-2 consistent way of talking about the narrative of 3
safety and showing the margin that we have in design 4
safety. And it retains a significant amount of 5
margin, too, and it shows you where that margin is.
6 Right?
7 Its not new but its a new way of talking 8
about it. Hence, the reason for this.
9 MEMBER MARTIN: All right. It sounds like 10 the key is the predicate of the SARRDL is to meet the 11 fuel limits. And so, that gets you basically 12 unchanged from existing Sodium Fast Reactors or even 13 Light-Water Reactors. And by whatever technology, 14 high temperature gas we had to do it a little 15 different.
16 MEMBER PETTI: Yeah. I think thats in my 17 mind the difference because there are no SARRDLs 18 because its difficult in that system to take because 19 its not a clad, pin and clad system.
20 So, basically, you know, even though the 21 SARRDL is there and its to demonstrate margin against 22 top 20, theres even more margin when one looks at the 23 design limits that you have that youre going to say 24 zero.
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51 MR. WILLIAMS: Uh-huh.
1 MEMBER PETTI: Or some very low number 2
But were going to analyze it up here with 3
SARRDLs and it will be higher, theres even more 4
numbers.
5 MR. WILLIAMS: Right. Exactly.
6 MEMBER PETTI: And once you get beyond AOO 7
space and its a, its a calculation based on fuel 8
performance leading to the larger source term, if you 9
will.
10 MR. WILLIAMS: Right.
11 MEMBER PETTI: Okay.
12 MR. WILLIAMS: Yep.
13 MEMBER ROBERTS: Is that different than 14 other reactor types? That sounds like the same thing 15 with Light-Water Reactors as when you get into, say, 16 LOCA space youve got to go calculate what their 17 relief fractions are or bound it conservatively.
18 It sounds are you doing anything different 19 there?
20 MR. WILLIAMS: I think, I think the 21 difference is in there youre going to have a lot of 22 additional features coming into play through the 23 mechanistic source term analysis than a traditional 24 Light-Water Reactor would have. I think more recent 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 Light-Water Reactor applications have looked at, you 1
know, fission product deposition and containment, 2
crediting additional things like that.
3 Were doing something like that, plus a 4
lot more because of the Sodium Fast Reactor features 5
and the lack of the pressurized system.
6 MEMBER PETTI: I think if you went back 7
in, you know, in the olden days, you know, with the 8
TID source term, that was just sort of an analysis.
9 But the 1.183 that we just looked at really is a 10 culmination of LWR source term that has more 11 mechanistic stuff behind it.
12 MR. WILLIAMS: Right.
13 MEMBER PETTI: But the reactivity events 14 do this, the LOCAs do this, and it was a way to 15 capture all of that. So, youre basically kind of 16 doing the same thing with the SARRDLs, the technology 17 is the same.
18 MR. WILLIAMS: Yeah, yeah.
19 MEMBER ROBERTS: Right. More mechanistic.
20 And just taking the worst of all accidents and finding 21 them.
22 MEMBER PETTI: Right.
23 MEMBER ROBERTS: But, fundamentally its, 24 its you go run your analysis, figure out what 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
53 release is because you violated the fuel limits 1
because of the nature of the accident. And then you 2
figure out what that is and you proceed with a 3
calculation.
4 MR. WILLIAMS: Yeah. We are still 5
demonstrating that we need the functional containment 6
performance criteria with an assumed major accident.
7 So, we are still doing that even though its probably 8
in the -- beyond the cutoff frequency in the PRA.
9 So, we are still taking that step. But we 10 are crediting all of the design features that we have 11 that are technology specific.
12 MEMBER ROBERTS: Sure. And thats 13 consistent with Sodium Fast Reactors in the past; 14 right? They looked at the protected loss of flow, and 15 your protected transient input powers, and those are 16 probably extremely low in frequency space.
17 MR. WILLIAMS: Right.
18 MEMBER ROBERTS: Im sorry. I guess 19 youll get to that, your plan is to look at 20 unprotected?
21 MR. WILLIAMS: Yes.
22 MEMBER ROBERTS: Okay, great. Thanks.
23 MR. WILLIAMS: Yes. Yeah.
24 MEMBER MARTIN: So, Eric, when you mention 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 that, you know, your analysis shows you this large 1
margin but with the, you know, mechanistic solution, 2
of course weve not seen that. Right?
3 MR. WILLIAMS: Right.
4 MEMBER MARTIN: And Im not sure where it 5
is in submittal and review space. You know, we can 6
only judge on what weve seen; right?
7 And traditionally DBCs have reason to fall 8
back on. But the idea is that it comes first and, you 9
know, and justification comes later, typically.
10 Now, in contrast, like the HTGR, you know, 11 they had the advantage of all the, all the testing 12 that was done at Idaho and kind of coincident with the 13 writing of the Reg Guide. Of course, there was a, you 14 know, fair amount of knowledge about how well TRISA 15 performed. And that insight kind of fed the writing 16 of that, of the Reg Guides. Im tracking it a little 17 bit.
18 But, you know, in other conversations Ive 19 had my understanding is that it influenced how, how 20 that was written. With the SFRs you dont have enough 21 read out there doing all this wonderful work for you.
22 You dont have that kind of in the, in the, you know, 23 public domain or, you know, working through it or see.
24 Its all just to say justification hasnt come to us, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 and so it makes it very difficult for us to depart 1
from what the Reg Guide says.
2 You know, and that was kind of the 3
criticism that came up during the subcommittee. So, 4
thats our, our perspective.
5 MR. WILLIAMS: We have some -- I mean, 6
Ill point you to the Argonne National Lab did a trial 7
mechanistic source term project in the ANL-ART series 8
of documents. RT-3 is essentially a PIRT on 9
mechanistic source term for Sodium Fast Reactor with 10 metal fuel.
11 TerraPower participated with them in that, 12 as well as other, other vendors. And were heavily 13 leveraging that work.
14 In fact, it talks a lot about the 15 fundamental safety of metal fuel under, you know, a 16 sub-cooled pool of sodium, and how it behaves, and the 17 retention of fission products within the fuel matrix, 18 the retention of fission products in the liquid 19 sodium, and how all of that behaves. And adds a 20 tremendous level of margin to safety.
21 And so, thats something in the public 22 domain that I think is really good background to read.
23 And then I think all of this will come together in the 24 mechanistic source term topical report. So, that I 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 think is coming up, so, can have a lot more discussion 1
about that.
2 MEMBER MARTIN: I think you might get a 3
different response if that came first.
4 MR. WILLIAMS: Yeah.
5 MEMBER MARTIN: Right? And it goes 6
through the process, the sausage-making.
7 MR. WILLIAMS: Yeah.
8 MEMBER MARTIN: But in contrast, you 9
brought this first. And so, it looks like the Reg 10 Guide strictly applies without any further 11 justification.
12 MR. WILLIAMS: Yeah. I can appreciate 13 that.
14 MEMBER PETTI: I said it in subcommittee, 15 you're not the first where because things are done 16 sequentially, we're trying to see the whole elephant 17 and all we see is a piece.
18 With other applicants, it wasn't until we 19 got to the PSAR, where there's numbers in there and I 20 went, oh. And the lightbulb goes off because you can 21 finally see all the pieces come together.
22 And it's like, God, I ask all these stupid 23 questions. I wish I knew this number when we started.
24 It'd be more efficient in the overall process.
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57 When we did this in NGMP, we asked a 1
source to -- it was a white paper at the time -- and 2
the fuel quality be done together, for exactly this 3
chicken-and-egg problem, when you're dealing with new 4
technology that's about a different education by --
5 it's just inherent in the new technology, I think.
6 MR. WILLIAMS: Yeah, these things used to 7
be historically very separate conversations you could 8
have. They were disjointed. And now, they're coming 9
together with LMP, which is a benefit. But we have to 10 have these integrated discussions a little bit better 11 now.
12 All right, I'll talk a little bit about --
13 MEMBER ROBERTS: Probably leads to what 14 you're about to say. One big-picture question comes 15 up with the Reg Guide at 1.232 and the fifteen-16 containment criteria, that resulted from decades of 17 progression -- if you look back at history, I'm sure 18 you're versed in all of this -- but going back in time 19 to S-PRISM, PRISM, IFR, go back, and all the different 20 developments in studying past reactors, there was a 21 1993 proposal to ease up on some of the requirements, 22 and you could argue that the results of that led to a 23 risk-informed, performance-based, set of containment 24 criteria for an SFR.
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58 And that's all got codified in Reg 1
Guide 1.232. So, it's kind of hard to see the case 2
for diversion from that given all the history, and the 3
fact that your client looks a lot like PRISM.
4 So, the accumulated judgment of all those 5
different generations of designers, led to those 6
criteria. So, maybe you can lead into your discussion 7
here with that as the starting point, is that when you 8
make what might be radical changes -- and that's my 9
question would be, are they really radical changes, 10 because I'm not quite sure they are -- but what appear 11 to be radical changes, then the justification overall 12 for what's different now, than what was in the 13 previous designers' minds, all those generations of 14 SFR developments, is kind of a question.
15 The second question is are you really 16 being different, because there are aspects of your 17 design that look a lot like the containment approach 18 for PRISM.
19 And so, it seems like you're actually 20 including the structures that PRISM had to meet their 21 containment objectives. And so, if that's the case, 22 then what's the implication of changing the criteria?
23 MR. WILLIAMS: Okay. Yeah, that's a great 24 segue. I was going to talk through some of the design 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 features that are different. There's essentially five 1
areas that I wanted to bring up and just talk through 2
a little bit.
3 And some of these are radical departures 4
from historical sodium-fast reactor designs. I'll try 5
and highlight those in particularly.
6 But first of all, this is not a departure.
7 But just the use of sodium coolant, of course, is just 8
a huge benefit in terms of having a low-pressure 9
system that doesn't have the forcing function for 10 radionuclide releases through the functional 11 containment, the maintenance of highly sub-cooled 12 sodium within the reactor vessel to retain fission 13 products -- also very important, so that's a part of 14 all sodium-fast reactors.
15 The fact that it's a full metal-fueled 16 core is also a departure. I think we are the first 17 fully metal-fueled cork. So, that takes away a couple 18 of things that were being looked at from the 19 hypothetical standpoint, that involved core 20 disruption-type accidents.
21 So, oxide fuel behaves very differently in 22 severe accident space than metal fuel. And of course, 23 having an integrated reactor vessel with a large pool 24 of sodium also makes the bulk boiling of the sodium 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
60 incredible event.
1 So, that, the maintenance of a highly 2
coolable geometry for metal fuel, that's one of the 3
departures.
4 The metal fuel also retains the fission 5
products in fuel matrix -- certain categories of the 6
fission produces, I should say.
7 And then having a pool reactor is another 8
big one. Because it drastically reduces the amount of 9
sodium piping that you have for the potential of 10 sodium fire.
11 So, keeping all of the primary system 12 piping inside the integrated reactor vessel, you've 13 removed the fundamental hazard, which is the best 14 thing you can do.
15 Then, what you're left with is the sodium 16 processing system, which is small-bore piping that is 17 contained within the functional containment boundary.
18 And I'll explain in a little bit how we address sodium 19 fires in a different way there too.
20 Because the bulk boiling is not a credible 21 event, what we are looking at is the potential of gas 22 bubbles in the fuel channel from fission product 23 release from a failed fuel pin. So, we have to look 24 at that.
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61 We're looking at potential entrainment of 1
cover gas as a way to get bubbles into the core. I 2
think that's highly unlikely to happen in such a large 3
pool reactor, but we are addressing that as well.
4 And when we look at that in our analysis, 5
we really don't see anything that would propagate fuel 6
pin failures within the assembly.
7 So, when you have a fission gas bubble 8
going up through the channel, the temperature of the 9
neighboring fuel pins barely increases. So, that is 10 looking to be a really good analysis.
11 So, we're not seeing any way for voids to 12 cause a large energetic release. So, that's all 13 coming from the fact that there's a pool reactor here.
14 And then one of the really big departures 15 is that there's no longer a sodium water steam 16 generator. So, that was one of the huge sources of 17 energetic release that pressure-retaining containments 18 had to address.
19 And so, by having molten salt energy 20 storage and having an intermediate sodium system, 21 we've eliminated that hazard from the design. So, 22 that's a key one.
23 We're also excluding any kind of water 24 suppression systems from inside the functional 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 containment, and being very careful not to introduce 1
water even in that space. So, addressing it there.
2 And then in terms of sodium fires, a lot 3
of other designs have used what appears to be similar 4
design features -- guard pipes, and things like that, 5
around sodium pipes -- but we have addressed that 6
entirely for the functional containment space, by 7
having a secondary barrier around all sodium piping 8
within the functional containment barrier, even the 9
intermediate sodium piping that's in that space above 10 the reactor. So, we're addressing that.
11 And then we have the guard vessel that 12 surrounds the reactor vessel. So, even an unlikely 13 leak from the reactor vessel would be contained in an 14 iterative space there.
15 So, by addressing these features in the 16 design, we've essentially eliminated those large 17 energy releases from the functional containment. And 18 as it talks about in the SECY paper, we really have 19 all of those conditions that would make a functional 20 containing approach fit.
21 We have a new coolant, we have a new 22 operating state, a close-to-atmospheric pressure, and 23 we've removed a lot of the major accidents, like 24 seeing generator and sodium water interactions.
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63 So, that kind of covers the design 1
features. And then talking a little bit more about 2
how the process looks when you're designing the 3
functional containment and setting the performance 4
criteria to take all these things into account, like 5
I said, it's really integrated together with LMP and 6
with SARDLs.
7 Because, essentially, the LMP is used to 8
establish the LBE categories that you're looking at.
9 And then the functional containment performance 10 criteria is established for each of those LBE 11 categories.
12 So, that includes the SARDLs. The SARDLs 13 are included for the normal and AOOs, 10 CFR 5034 for 14 the d/b/a's, etc.
15 And then what you do is you establish 16 clear barrier performance criteria for all the SSEs in 17 the functional containment. So, again, that goes back 18 to the design process and integrating the safety 19 requirements with the design from the beginning, and 20 being able to review those over and over again as you 21 go through the LMP process.
22 And then we do demonstrate that those 23 performance criteria met with a major accident. And 24 if you want to reference back to the July 2023 meeting 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
64 we had with the NRC staff, they addressed SARDLs, 1
functional containment, and the major accident, and 2
kind of laid out the different major accidents we 3
would be using, and some of the main assumptions going 4
into those.
5 And then finally, the mechanistic source 6
term then pulls from all of these areas, to actually 7
go through and demonstrate the safety margin.
8 And then what you see, like I said before, 9
the specific PRA results on a frequency consequence 10 curve.
11 So, that's how functional containment 12 works with all the other elements of the LMP.
13 MEMBER ROBERTS: I have two questions of 14 what you just now laid out. Of all the features, it 15 seemed like all of them are also characteristic of 16 PRISM, except for the steam generator they moved 17 outside of the containment.
18 So, is there enough of a difference from 19 PRISM that the thought process that went into the 20 PRISM containment approach is no longer needed?
21 MR. WILLIAMS: In my opinion, yes. I 22 think we've also gone further with the sodium fire 23 protection within the functional containment space, 24 than was done in PRISM. I'd have to go back and check 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 that.
1 But the steam generator removal is a 2
pretty huge one, I think. And I think applying it 3
with LMP is also different, because the LMP gives you 4
the hooks, if you will, to demonstrate functional 5
containment, along with all the other safety features, 6
throughout the design process.
7 MEMBER ROBERTS: If you look at the 8
approach PRISM used to containment, it was a guard 9
vessel, and I guess they call it the containment dome, 10 which is the equivalent of your -- area access 11 enclosures, as I read it.
12 And looking at the PSAR that staff just 13 accepted a couple of weeks ago, one of the criteria 14 that you list is maintain at least one barrier between 15 the clotting, piping, or vessel, containment-ready 16 nuclide source to withstand all the design basis 17 access conditions, and whose leakage is specified by 18 design requirements for testing, which sounds a lot 19 like the PDCs are going to take them out.
20 To have a containment structure, to have 21 leak testing, design requirements that ensure that the 22 leakage rates are met, that seems a lot like the 23 containment requirements that are in the SFR design 24 criteria.
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66 Again, that's not to say -- it's just a 1
very radical change. You talked about having some 2
sort of containment around the intermediate sodium 3
piping. Is that the head area access, or is that 4
something else?
5 MR. WILLIAMS: Yeah, essentially, for the 6
core, the primary safety-related boundary essentially 7
is the reactor coolant boundary. And then the 8
secondary barrier is essentially the guard vessel and 9
the head access area combined.
10 So, those are the boundaries that you 11 think of for functional containment. And so, 12 essentially, what we would do is meet that criteria 13 through the GDZs on the primary coolant boundary, that 14 are essentially very similar to that.
15 MEMBER ROBERTS: All right. So, as it 16 seems almost like rearranging the deck chairs, that 17 either you have the same containment capability, it 18 seems like that S-PRISM, PRISM, the plants that were 19 the foundation of the SFR design criteria, you have 20 that, you're going to have to have design criteria to 21 show you meet them, which it seems like those are the 22 design criteria that are specified for SFR design 23 criteria in the appendix of the Reg Guide.
24 So, again, I'm just trying to understand, 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 are you really doing something radical, or is this 1
really just making the terminology map up with LMP?
2 MR. WILLIAMS: I don't think the head 3
access area is a pressure-retaining containment. So, 4
the fact that we don't have the sodium water reaction 5
in that space to require that is probably the main 6
difference there.
7 MEMBER ROBERTS: Right. And if I remember 8
at the SFR design criteria, it allows you to figure 9
out what relatively low pressure you would need for 10 that pressure-retaining containment.
11 So, that's already one of the performance-12 based allowances in the Reg Guide. Again, maybe it 13 requires some more thought, but it seems like what 14 you're doing isn't necessarily, from a design 15 perspective, much of a change from PRISM and what the 16 SFR design criteria are trying to push.
17 In which case, maybe you'll end up putting 18 them back in. I don't know, I'm just trying to 19 understand. But seems like you would need testing 20 requirements for leakage if you have a requirement, 21 self-imposed, that you have leakage specified by the 22 time requirements for testing.
23 MR. WILLIAMS: Yeah, I assume we would be 24 meeting all of those requirements and demonstrating 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 that anyway, even without the criteria in there.
1 MEMBER ROBERTS: Okay. Yeah, I think that 2
makes sense. Thanks.
3 MEMBER PETTI: I just think LMP provides 4
a structure.
5 MR. WILLIAMS: Yeah. Mm-hmm.
6 MEMBER PETTI: That if you develop these 7
design criteria, let's say from the bottom up and 8
years of experience, LMP gives you kind of a top-down 9
way to look at it and make sure you meet in the 10 middle.
11 And we kind of assure you there's 12 designers early in the process, that the requirements 13 have the right, the requirements at a broad system 14 level, that in principle, going bottoms-up you could 15 miss something. Right? And then go, oh yeah, down 16 here we got to go backtrack.
17 LMP, if it's done iteratively, like it 18 says, prevents or minimizes that sort of backtracking.
19 MR. WILLIAMS: Yeah. Yeah, that's right.
20 MEMBER PETTI: Yeah.
21 MR. WILLIAMS: That's right. I think it 22 makes the conversation clearer, and I think there's a 23 lot of value in that.
24 MEMBER ROBERTS: Did we already cover what 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 you planned to present?
1 MR. WILLIAMS: Yes, yes, we did, I think.
2 MEMBER ROBERTS: Okay, great.
3 MR. WILLIAMS: Yeah, we did.
4 MEMBER ROBERTS: Very helpful.
5 CHAIR KIRCHNER: Could you just flesh out 6
for us an example -- let's pick on something that you 7
already identified as one of your design features and 8
one of your barriers. What would the performance 9
criteria look like for the guard vessels?
10 MR. WILLIAMS: Yes. So, there's a couple 11 of key criteria on the guard vessels. So, 12 essentially, a postulated leak from the reactor vessel 13 has to be contained within the guard vessel, and the 14 gap within the guard vessel is a size such that it 15 remains above the heat exchangers and the reactor 16 vessel and the pumps so you can continue to provide 17 aquicore cooling.
18 Yeah, it has a function there. It also 19 carriers a radionuclide retention function as a 20 secondary barrier for the functional containment, the 21 primary barrier being the reactor vessel.
22 So, if you assume the fuel pins have 23 failed -- and we assume all the fuel pins have 24 failed -- and demonstrating this, we assume the failed 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 barrier there, if you also have a failure in the 1
reactor vessel barrier, then the guard vessel is there 2
to prevent further leakage.
3 CHAIR KIRCHNER: So, effectively, when you 4
implement these requirements, you're going to have 5
something that -- don't like to use the LWR 6
terminology -- and essentially leaked barrier about, 7
if not even a more demanding requirement, regardless 8
of the fact that you have sodium --
9 MR. WILLIAMS: In some cases it is more 10 demanding, because we're trying to prevent the sodium 11 from contacting air as well. Yeah.
12 CHAIR KIRCHNER: Members?
13 MEMBER ROBERTS: Okay, it sounds like 14 we've no more questions. So, thank you very much for 15 your presentation, TerraPower, and I guess we'll 16 switch to the NRC staff now.
17 (Off-mic comments.)
18 MEMBER ROBERTS: Then let's take a break 19 until ten o'clock, giving the staff a chance to set 20 up.
21 CHAIR KIRCHNER: So, for those online 22 listening in, we're going to take a break until ten 23 o'clock, Eastern Time.
24 (Whereupon, the above-entitled matter went 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 off the record at 9:50 a.m. and resumed at 10:01 a.m.)
1 CHAIR KIRCHNER: Okay, we're back in 2
session. And I just want to go ahead and turn it back 3
to Tom. Go ahead, introduce the NRC.
4 MEMBER ROBERTS: Thank you all. And I'm 5
just going to go ahead and pass it over to the NRC 6
staff. Mallecia, are you going to start? Or 7
Stephanie?
8 PARTICIPANT: Actually, we're going to 9
have Reed start it.
10 MEMBER ROBERTS: Reed start it. All 11 right.
12 MR. ANZALONE: I'm just going to take it 13 from the beginning.
14 MEMBER ROBERTS: All right, go ahead. Go 15 ahead, Reed. Thanks.
16 MR. ANZALONE: Thank you, Member Roberts.
17 So, I will jump straight into it. We had most of the 18 members for the subcommittee meeting. So, I think the 19 goal is just to try to cover the key points from that 20 subcommittee meeting. And I think TerraPower did a 21 good job of laying out a lot of the technical aspects 22 related to PDC. So, we're going to basically just 23 focus on our approach for the review.
24 So, I'll briefly cover the purpose of the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 topical report and our strategy for the review, talk 1
a little bit about the regulatory requirements, give 2
a real brief overview of the PDCs, and then jump into 3
key topics from the subcommittee meeting, which are 4
functional containment SARDLs and the limitations and 5
conditions. And the slides aren't advancing.
6 Okay. So, the purpose of the topical, 7
like TerraPower talked about, was to describe the 8
process for developing PDCs, and then actually give us 9
those PDCs. And that's partially to address 10 compliance with 10 CFR 5034.
11 They also wanted to describe their 12 rationale for meeting the intent of PDC 26. I'm 13 actually not going to talk about that today, just to 14 be clear, to focus on the topics from the subcommittee 15 meeting.
16 And then our strategy for the review was 17 to review the PDC's conformance with the Reg Guide 18 group and evaluate the deviations from the Reg Guide, 19 considering the key design features.
20 And really, our scope, we wanted to -- and 21 this is something that we struggle with a little bit 22 as the staff for PDCs, because it's very easy to get 23 into the technical details of how you're going to 24 comply with the PDCs -- we wanted to focus on whether 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 the PDCs themselves were acceptable.
1 And the design is an appropriate and 2
necessary context for that, but we didn't want to get 3
too into the weeds on how they were going to meet 4
them.
5 Part of our review then also was 6
identifying the interaction between the Reg 7
Guide 1.232 approach and the LMP, which TerraPower 8
talked about a little bit today. And then the PDC 26 9
rationale was a specific subject that we tackled, 10 that, again, I'm not going to talk about really today.
11 Apparently, I don't know how to move the slides 12 forward.
13 All right. Okay. So, the regulations, I 14 already mentioned that 5034 requires the CP applicant 15 to include the PDCs. TerraPower had the topical 16 report, which I believe is incorporated by reference 17 in the PSAR, but then they also put the PDCs into the 18 PSAR as well. But this was submitted well in advance 19 of the construction permit application.
20 And then Part 50, Appendix A, which has 21 the general design criteria, provides requirements on 22 the scope and content of PDCs, for all reactors, 23 including non-light water reactors. So, that first 24 bullet there talks about what the PDCs need to be able 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 to do.
1 And then the second bullet is really sort 2
of more guidance saying that the GDCs that are in 3
Appendix A provide guidance for how the PDCs should 4
look.
5 So, TerraPower developed their PDCs, like 6
they mentioned, based on Reg Guide 1.232. Most of 7
their PDCs were directly based on the SFRDC, which 8
were in Appendix B of that Reg Guide.
9 Some PDCs were based on the modular high-10 temperature gas reactor design criteria, which are in 11 Appendix C, and those were generally used to implement 12 functional containment, or reflect the use of SARDLs.
13 Most of the PDCs were modified in one way 14 or another from the base design criteria in the Reg 15 Guide, and we kind of circled around it a little bit 16 in the conversation earlier with TerraPower.
17 But there are no design criteria for those 18 numbers down there, due to the use of functional 19 containment. Talk about that a little bit in the next 20 couple of slides.
21 So, these were the general changes to PDCs 22 that I laid it out in subcommittee meeting. We're 23 going to focus on those two today, just for the sake 24 of keeping things a little tighter.
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75 So, starting with functional containment, 1
so, obviously, discussions on containment -- I think 2
members have mentioned this earlier in the meeting --
3 the discussions about containment and functional 4
containment, and what's appropriate for NSFR 5
containment, have been going on for a long time.
6 Part of that discussion is Reg 7
Guide 1.232, which has containment criteria in it for 8
SFRs and for MHTGRs.
9 But one thing I will say is that Reg 10 Guide 1.232 came out, and then SECY 1896 came out, 11 which actually sort of codified the functional 12 containment approach.
13 And so, I'll talk about that more on the 14 next slide, our take on that SECY paper and the 15 associated SRM.
16 But sort of even at a high level, Reg 17 Guide 1.232 talks about functional containment. Yes, 18 it's in the MHTGR DC, and I think a lot of the impetus 19 for developing that concept came from the HTGR world 20 and TRISO fuel.
21 But the approach is technology-inclusive.
22 And the Reg Guide says it's applicable to advanced 23 non-light water reactors without a pressure-retaining 24 containment structure.
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76 So, our thinking -- and I'll go to the 1
next slide to talk a little bit about the SECY 2
paper -- is that from the get go as part of the 3
current conversation on functional containment, the 4
idea is that it is technology-inclusive, risk-5
- informed, and performance-based approach to 6
containment design criteria.
7 The SECY paper, which was approved by the 8
Commission, gives a methodology for determining 9
functional containment performance. That developed 10 into LMP, and it was developed in parallel with the 11 Reg Guide, which noted that some of the stuff still 12 needed to be approved by the Commission.
13 MEMBER ROBERTS: Maybe you could comment 14 on -- I'm going to make an assertion and you can tell 15 me where I'm wrong.
16 It seems like the Appendix B -- SFR design 17 criteria for containment -- are technology-inclusive 18 for an SFR, risk-informed, performance-based, because 19 if you look at the history, it seems like the NRC took 20 a turn at that probably 30 years ago, and said, we 21 need to go revise the GDC that are derived from light 22 water reactors, because they don't really apply to 23 this technology, and what's left there does have some 24 aspects of at least performance-based and, I'd expect, 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 risk-informed.
1 So, is it fair to say that the existing 2
SFR design criteria for containment are risk-informed, 3
performance-based?
4 MR. ANZALONE: I would say that that's 5
true to a degree. And I think I -- and maybe it's the 6
next slide -- I'm going to talk a little bit more 7
about the specific SFR design criteria, which do have 8
kind of notes in them about, this would apply under 9
these certain situations.
10 But at the same time, if you go back and 11 look at the SECY paper, I had the benefit of going 12 through the transcripts from the ACRS meeting and your 13 letter on this, and I think sort of conceptually, the 14 thing that functional containment as an approach does, 15 is it's capable of encapsulating all of the possible 16 different approaches.
17 So, if you look at SECY, I think it's 18 93092, which might have been what you were talking 19 about 30 years ago.
20 There's a bunch of different containment 21 designs that are referenced in that. There's the 22 MHTGR, which is sort of a pure functional containment 23 along these lines.
24 There's the PRISM reactor, and I think it 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 was the OR S design. That all have kind of varying 1
degrees of containment, leak tightness, and different 2
containment designs.
3 And if you look back at the SECY paper and 4
sort of -- a bunch of the discussion around that, the 5
idea was that functional containment performance could 6
be, you could define a generic functional containment 7
criteria that could encapsulate all of that.
8 So, I think the approach that was in 9
SEC 1896, is intended to kind of wrap around all of 10 them. And I think the letter that the ACRS wrote at 11 the time actually kind of explicitly says, hey, maybe 12 the staff should go back and revise Reg Guide 1.232 to 13 say, hey, this concept could apply across the board.
14 So, that was part of our consideration 15 here in thinking through does functional containment 16 make sense for TerraPower?
17 CHAIR KIRCHNER: Reed, let me help you 18 here. I have the letter.
19 (Laughter.)
20 CHAIR KIRCHNER: And it goes on to say 21 that the containment criteria in Appendices A, B, and 22 C of the draft Reg Guide are logically inconsistent.
23 So yes, there was this thought that they should be 24 technology-inclusive.
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79 So it wasnt clear at the time whether the 1
staff would go back and revisit the Reg Guide. I 2
think there was a pointer there that the sets of 3
criteria werent, as we stated, logically consistent.
4 MEMBER ROBERTS: Yeah, and were 30 years 5
later and this is Natrium, not PRISM, and theres a 6
lot of development of risk-informed thought processes.
7 And that can certainly result in another term.
8 But it seemed to me, Id want to see, get 9
your reaction, that the existing Reg Guide Appendix B 10 is risk-informed, performance-based for that specific 11 technology. And that doesnt mean thats set in 12 stone, because there are changes to, you know, from 13 PRISM to Natrium, and there are changes in thought 14 processes, or risk-informed space.
15 And it seems like okay, an unfair 16 statement to say this is the addition of a functional 17 containment thought process, because you could argue, 18 you know, that term wasnt used. This was essentially 19 a functional containment for and SFR developed 30 20 years ago, just not using that term. Is that fair?
21 MR. ANZALONE: Yeah, and I can -- I just 22 moved on to the next slide, because I think this sort 23 of talks about what youre getting at. There are 24 certain of the SFRDC that talk about, you know, how 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
80 they would be applicable if certain approaches to 1
containment were taken or might not be -- like maybe 2
you would have the SFRDC, but it wouldnt actually be 3
applicable to any structures at the plant, which is 4
kind of an odd thing.
5 So thats the, basically I think its, 6
yeah, 39 -- 38, 39, 40, and 50-57 are all sort of in 7
that space where, you know, you could maybe make the 8
argument, okay, we dont need this structure, even 9
though we have these criteria. But it, to me, its 10 not, thats not like a clean approach. Thats messy.
11 And I understand that, you know, the SFR 12 or the functional containment criteria is itself a 13 little bit messy because we kind of get everything at 14 once with the demonstration of functional containment 15 performance. But the criterion itself is more 16 straightforward, and you dont have to make these 17 arguments about how we have these criteria but theyre 18 not actually applicable.
19 So like for example, if they didnt need 20 heat removal in the containment, then they wouldnt do 21 anything with 38, 39, or 40. But then why do you have 22 them at all? And to me, it makes more sense to apply 23 like a sort of more straightforward performance-based 24 criterion that encapsulates everything.
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81 So our take is that its, you know, 1
appropriate to apply the functional containment 2
criterion for TerraPower. And Ill, I guess Ill go 3
onto the next slide.
4 Because you know, the SECY paper talks 5
about it being acceptable for non-light water 6
reactors. We do think, and TerraPower I think laid 7
these out very well, there are attributes of the 8
reactor design that are necessary to be able to, you 9
- know, effectively actually use a
functional 10 containment approach.
11 But that functional containment 12 performance still needs to be demonstrated, and thats 13 part of our review in the construction permit 14 application. And you know, based on what I read in 15 the transcripts and the discussion surrounding this 16 issue back when ACRS reviewed it back in 2018, there 17 were a few sort of thoughts about, you know, defense-18 in-depth and how you would go about actually analyzing 19 those.
20 So I just wanted to throw these points in 21 here that LMP is really like a key part of this. And 22 its part of why we have that limitation condition 23 that says thou shall use LMP is youre going to apply 24 this approach. It implies that youre going to have 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 a PRA and a mechanistic source term. It gives you 1
criteria that you need to meet.
2 You have to explicitly consider 3
uncertainties. And you have to do this risk-informed, 4
performance-based, defense-in-depth adequacy 5
evaluation thats in NEI 18-04. So you know, 6
functional containment doesnt mean no containment, it 7
means you evaluate all of the barriers that are in the 8
way of the release of radionuclides.
9 So if we were going through TerraPowers 10 evaluation in our review and we came across something 11 that we felt like releases werent being appropriately 12 addressed, thats something that we would bring up 13 during our review.
14 Okay, any questions? Because Ill be 15 moving on to SARDLs, which thats a pretty brief 16 discussion.
17 So SARDLs were initially identified for 18 TRISO fuel for the MHTGR. Theyre for normal 19 operations in AOOs, and they need to be established so 20 the Part 20 limits arent exceeded. But the SECY 21 paper on functional containment performance criteria 22 does pretty much say SARDLs are intertwined with 23 functional containment performance criteria.
24 And so, you know, I think the concept 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
83 we understood from talking with TerraPower about this 1
is that the impetus for SARDLs is that use of 2
functional containment, which is also all intertwined 3
with LMP.
4 And so that first bullet here on this 5
slide is that we -- our view is that SARDLs are 6
appropriate (audio interference) and consistent with 7
a performance-based evaluation.
8 We already talked about fuel design limits 9
that can be used to help evaluate those SARDLs. And 10 Im really glad that Eric touched on it during his 11 presentation. I think one of the key things is that 12 SARDLs are a useful tool for looking at ex-vessel 13 events and sources of radionuclides other than just 14 the fuel inside the reactor.
15 And he mentioned that ANL art series of 16 reports looking at mechanistic source terms. One of 17 the things that ANL has found, and I think 18 TerraPowers assessment also agrees with this, the 19 things that drive the plant risk are not the in-vessel 20 events. Theyre all of the issues in these like 21 auxiliary systems and fuel-handling accidents.
22 So having SARDLs to evaluate those events 23 actually helps a lot. Part of SARDLs is that you 24 would need to include a means of monitoring activity 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 in these systems, so thats something that well 1
evaluate as we look at their plant design.
2 And the SARDLs need to still be proposed 3
and evaluated, and we did discuss them. Eric 4
mentioned the July public meeting, where we had some 5
example SARDLs that we talked about with them.
6 Moving on to the scope and applicability 7
of PDCs, and really this I just wanted to say, you 8
know, we talked about possible changes to the 9
limitation 2 and RSE. As of right now, we havent 10 identified any changes, and so we didnt pass along to 11 the ACRS. So thats why I wanted to bring that up 12 again in this meeting. So that limitation 2 really is 13 focused on the use of LMP.
14 And these are the same conclusions from 15 the subcommittee meeting, so I wont reiterate them in 16 the interest of trying to get us closer to the 17 schedule in the agenda, so. Anyway, happy to take any 18 questions that you may have. If not, then we can --
19 CHAIR KIRCHNER: Reed, when you went 20 through this, okay, so you accept the premise. Did 21 you systematically look at the implications of 22 expunging, or maybe a better way to say it is to 23 divert from the ensuing GDCs that are containment-24 related? See where Im going with this?
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85 If you say 16 is now functional 1
containment rather than containment, then do you still 2
systematically look at all those other GDCs that 3
support containment? Because what they by and large 4
do is protect that fission product barrier in once 5
sense or other. So was that --
6 MR. ANZALONE: So yeah, there a couple --
7 CHAIR KIRCHNER: What youre thinking?
8 MR. ANZALONE: Yes, and there are a couple 9
of PDCs that TerraPower added back that talk about the 10 performance of the reactor building envelope and stuff 11 like that are -- that are necessary when you use a 12 functional containment approach.
13 CHAIR KIRCHNER: You feel that youve got 14 a complete set and that would address those other 15 functions that, how should I say this, that the 16 containment building structure provided, went beyond 17 just fission product release. Either they were 18 protecting the building de facto for a LWR becomes the 19 major protection against external events, or many 20 external events, etc.
21 So the containment function goes beyond 22 just fission product barrier purposes.
23 MR. ANZALONE: Yeah.
24 CHAIR KIRCHNER: Protecting against 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 internal, external events, flooding, fires. So when 1
the staff did its review, you felt there was a 2
complete set of those other functional attributes that 3
go into -- unfortunately, containment for an LWR is a 4
multipurpose --
5 MR. ANZALONE: Right.
6 CHAIR KIRCHNER: Function. So youre 7
satisfied that they address those -- their thing.
8 MR. ANZALONE: Yeah.
9 CHAIR KIRCHNER: So when we come back and 10 look at an actual detailed design and look at, let me 11 pick one of the things thats always problematical 12 with containment is double isolation valves, inside, 13 outside, and so on.
14 You still feel that the functional 15 equivalent of those containment-like criteria would 16 still be applied when you reviewed individual fission 17 product barriers, i.e., a guard vessel?
18 MR. ANZALONE: Yes.
19 CHAIR KIRCHNER: Think someone was just 20 unmuted there. Okay.
21 MR. ANZALONE: But yes.
22 CHAIR KIRCHNER: All right, thank you.
23 MEMBER ROBERTS: Yeah, I guess I thought 24 of it a little bit differently. Because 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
87 functional containment is not entirely defined yet, if 1
its defined as the PSAR indicates with containment 2
structures and leak test requirements, and the like, 3
you would add back in criteria as the design of the 4
containment would like more and more like a classic 5
SFR containment or light water reactor containment.
6 That, that was my interpretation as you 7
got the big picture, you know, this is a functional 8
containment, were going to figure out what it means.
9 But then when it looks like a more conventional 10 containment, you have to look at putting back in these 11 kind of design criteria.
12 Whether theyre called PDCs or what you 13 call them I dont know, but I would think youd still 14 want to make sure the requirements for leak testing or 15 the requirements for double valve isolation, whatever 16 they happen to be, are met once the containment 17 structure looks like a structure that these were 18 applied to.
19 MR. ANZALONE: Im not sure I followed.
20 MEMBER ROBERTS: Yeah, thats still 21 puzzling me a little bit. The proposed containment 22 for the PSAR is to have a structure, right. So there 23 is a structure around each boundary that contains 24 reactor material, which ends up looking a lot like, 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 you know, a PRISM containment.
1 And so it looks like a PRISM containment, 2
then the requirements that were based on PRISM 3
containment will seem to be met in some form or have 4
to be met.
5 And so whether you call those PDCs or call 6
them design requirements or tech specs or whatever 7
they happen to be, once you -- once they go back to 8
the design looks like PRISM, then the requirements are 9
imposed in PRISM structure would seem to be evaluated, 10 need to be evaluated for applicability and things like 11 leak test capability would seem to need to be a 12 requirement. Just like PDC-52.
13 MEMBER ROBERTS: So some of those detailed 14 design requirements would, you know, depending on the 15 specific design of the system would I would expect 16 sort of flow down from the high level performance 17 requirement in the PDCs.
18 CHAIR KIRCHNER: So where I was coming 19 from is if you were to go back and look at that -- and 20 Im sure you have -- the staffs work on functional 21 containment, they point to additional sets of 22 functional containment performance standards, like 23 protecting other risk-significant SSCs.
24 MR. ANZALONE: Yeah.
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89 CHAIR KIRCHNER:
Support of
- them, 1
occupational radiation exposure, removing heat, 2
physical protection, like security for external 3
events, etc. So thats where I was going.
4 MR. ANZALONE: Yeah, thats what I 5
understood from you. From Member Roberts, I thought 6
you were talking about sort of more detailed criteria 7
for specific system designs. Is that correct?
8 MEMBER ROBERTS: Yeah, Im thinking if 9
these 15 criteria that are in Appendix B or based on 10 the characteristics of a PRISM containment structure, 11 and if the nature of a containment structure looks 12 like a PRISM containment structure, then the same 15 13 requirements would seem to need to apply in some form.
14 Whether you call them derived requirements 15 or principal design requirements or whatever, if the 16 containment structure requires them to meet its 17 function, then they would need to be tracked I would 18 think in some form.
19 MR. ANZALONE: Well, I guess I would say 20 that with that high-level performance-based 21 requirement, you know, TerraPower would look at their 22 design and they would look at the releases and the 23 doses, and they would figure out which of those kinds 24 of criteria would need to be looked at.
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90 So if you look at like double isolation 1
across containment, I dont -- I dont think that 2
something like that is part of TerraPowers design.
3 I dont know for certain, and Im sure it would depend 4
on the specific system. So thats maybe an example of 5
where like using a performance-based criterion would 6
buy you something in design space.
7 MEMBER ROBERTS: And if just folks want an 8
example, Criterion 52 says a reactor containment 9
structure and other equipment that may be subjected to 10 containment test conditions shall be designed so the 11 periodic integrated leak rate testing can be conducted 12 to demonstrate resistance and containment design 13 pressure.
14 So if their design is going to have a low 15 leakage, you know, structure for the -- the head 16 access area, and their intent is to verify that by 17 test, then why wouldnt 52 apply?
18 MR. ANZALONE: Because its encompassed by 19 this functional containment performance criterion.
20 MEMBER ROBERTS: So it ends up being a 21 derived --
22 MR. ANZALONE: Yeah.
23 MEMBER ROBERTS: Not a principal design 24 requirement.
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91 MR. ANZALONE: I would agree with that.
1 MEMBER ROBERTS: Okay. But in some form, 2
it would seem like your review of that system, once 3
its concluded that it looks like PRISM --
4 MR. ANZALONE: We would --
5 MEMBER ROBERTS: You would --
6 MR. ANZALONE: We would want to make sure 7
that it met its performance requirements. That the 8
performance requirements made sense and that it met 9
them. But that is all down the road as part of our 10 construction permit application review.
11 VICE CHAIR HALNON: But so -- this is Greg.
12 Those performance requirements, I mean, to your point, 13 if its not required for part of the functional 14 containment definition of whatever that SSC, if you 15 would, then you wouldnt have to do the leak testing.
16 This really doesnt apply.
17 MR. ANZALONE: Right.
18 VICE CHAIR HALNON: Its not part of the 19 basis for that containment. I mean, so its not 20 really containment structure. Thats the way Im 21 reading that. I dont have a conflict here.
22 I see it -- see what -- the functional 23 containment, you could look at that as an SSC, even 24 though its distributed amongst things. This is not 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 part of that SSC. Thats the way Im looking at it.
1 So I dont see a conflict in my mind.
2 MEMBER ROBERTS: It depends what is in the 3
system to meet the functional containment 4
requirements.
5 VICE CHAIR HALNON: Right. So if that 6
vessel, guard vessel, you say is not supposed to be, 7
you know, a gaseous leak-tight, if you would. Its 8
sodium leak-tight. It doesnt make -- in my mind, 9
its just not part of that functional containment 10 requirement. So theres no -- 52 wouldnt apply.
11 Even though you say its derived, its sort of 12 derived. But its not part of the SSC for functional 13 containment.
14 Im not looking at specifically design, 15 Im looking at conflict. I understand how 52 would 16 not be part of this because its not part of the SSC 17 of functional containment, in a classic sense.
18 Anyway, I just wanted to make sure that I 19 understood why you said it was derived. No leak test 20 is required because its not part of the SSC.
21 MEMBER ROBERTS: Yeah, it depends on the 22 functional containment model on what theyve got in 23 there. Well, what they say in PSAR is to have these 24 structures surrounding the vessels and pipes and 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
93 like that contain radioactive material. And they plan 1
to leak-test everything thats in that.
2 So the question would be whether that 3
derived requirement ends up being something that the 4
NRC staff would validate because thats part of the 5
approach taken to containment that looks like PRISM, 6
or something thats just part of developing the 7
functional containment model and whether or not its 8
derived from that.
9 Im not sure that distinction is clear, 10 but if the structural containment looks like PRISM and 11 its credited, you know, similar to the way it was 12 operated in PRISM, then the requirements there were 13 applied to PRISM would seem to apply also.
14 VICE CHAIR HALNON: Unless theyre subsumed 15 into something bigger.
16 MEMBER ROBERTS: Yep.
17 Any other questions for Reed on this 18 subject?
19 CHAIR KIRCHNER: Reed, at some risk Im 20 going to bring up PDC-26. I didnt want you to get 21 off that easily. Can you just for the record, since 22 this is full committee, give us your evaluation of the 23 proposal for PDC-26?
24 MR. ANZALONE: Yeah, sure. So, and we did 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 cover this at the -- I dont think Im going to say 1
anything different than I said at the subcommittee.
2 CHAIR KIRCHNER: I dont expect that.
3 MR. ANZALONE: For the record, you know, 4
TerraPower proposed that they would essentially adopt 5
the SFRDC-26 with some conforming changes about safety 6
significance that are consistent with LMP that are 7
applied to all the different PDCs. So its 8
essentially, I would say its essentially unchanged 9
from SFRDC-26 in like a meaningful way.
10 And so then they proposed that they would 11 meet that by essentially showing they have two 12 different -- two different control rod designs that 13 mitigate common cause failures between the different 14 control rod designs.
15 And so they were intending to show that 16 there was sufficient diversity and independence 17 between the different control rod design. And theres 18 a different means of control rod insertion. So they 19 wanted to show that in a sort of risk-informed manner, 20 that that would be independent and diverse enough to 21 meet Criterion 26.
22 CHAIR KIRCHNER: But the staff position, 23 as I understand it now, is you basically accept that, 24 but its TBD --
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95 MR. ANZALONE: Exactly.
1 CHAIR KIRCHNER: D being demonstrated by 2
design that theyre not going to be subject to a 3
common cause failure.
4 MR. ANZALONE: Correct.
5 CHAIR KIRCHNER: Seismic misalignment such 6
that you cant insert either set of control rods. So 7
this, Im just flagging this because its a major 8
departure from what in the past had been the 9
definition of diverse, looking at two diverse or 10 similar systems for that function.
11 MR. ANZALONE: Yep. One thing I will say 12 fortunately that SFRs buy you, and I know this was 13 mentioned during the subcommittee meeting, you know, 14 you can fail a lot of control rods and still get 15 enough negative reactivity insertion to shut down the 16 reactor. So I think that would help with the overall 17 demonstration, that theres enough diversity there.
18 CHAIR KIRCHNER: Perhaps the exponents of 19 the opposite take would say with an SFR, you can get 20 a significant reactivity insertion event.
21 MR. ANZALONE: Absolutely, and thats 22 something were going to be really focused on in our 23 review.
24 CHAIR KIRCHNER: Yeah. Because the one 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
96 thing, if were here Ill just state this, is that 1
were now at a size for a fast reactor design like 2
this, that youre at the edge of where you can rely on 3
the leakage as a negative feedback mechanism. So I 4
expect that when we review the PSAR, that this will be 5
looked at very carefully when were considering the 6
PDC-26 as well.
7 MR. ANZALONE: Absolutely, absolutely, I 8
totally agree.
9 MEMBER ROBERTS: Any other questions from 10 the members or consultants on the PDC? Now we can 11 move on to the fuel qualification report.
12 MS. DE MESSIERES: Actually, this is 13 Candace de Messieres from the NRC.
14 So I just wanted to make one clarifying 15 point for the record as it relates to L&C No. 2, that 16 at the highest level, that L&C has to do with the 17 synergy between the frameworks between PDC and LMP.
18 And that the staff continues to work to ensure 19 clarification at a generic level on that issue.
20 So I just wanted to make that note for the 21 record. Thank you.
22 MEMBER ROBERTS: Thank you.
23 MR. ANZALONE: Okay, so moving on to fuel 24 and control assembly qualification. So we started 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 with fuel and well end with -- well let fuel take us 1
out.
2 This is a, I would say, more just 3
relatively straightforward truncation of my 4
presentation at the subcommittee, so I can go through 5
this as quickly or as slowly as we want. So Ill talk 6
a little bit about, again, the topical report purpose 7
and our strategy in the review. Ill talk a little 8
bit about the regulatory requirements in the guidance.
9 And part of what we used a lot in our 10 review was this NUREG-CR 7305 for giving us technical 11 information that we could use to help evaluate 12 TerraPowers fuel. Then Ill go through a brief 13 overview of our safety evaluation and the overall 14 conclusions.
15 So the purpose of the topical report was 16 to provide a plan to qualify Natrium Type 1 fuel, 17 which as TerraPower talked about, is a U-10Zirc 18 metallic fuel in HT9 cladding. And theyre control 19 assemblies. And it requested NRC review and approval 20 of a bunch of different items that essentially are the 21 fuel qualification plan.
22 And it provides some fuel qualification 23 results and talks about their ongoing plan of fuel 24 qualification activities.
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98 So our strategy in the review was to 1
review the scope and adequacy of the plan in the 2
context of NUREG-2246, which was released after 3
TerraPower had started developing this topical report.
4 So they included a crosswalk that sort of referenced 5
their criteria that they came up with against NUREG-6 2246.
7 And then we also reviewed it, as I 8
mentioned, against NUREG-CR 7305, which you know, I 9
should say is not, its not -- it doesnt have like 10 the status of guidance, right. Its not a reg guide.
11 But it is additional technical information that we had 12 contractors from several different national labs put 13 together to help us look at metallic fuel.
14 So the regulatory requirements. And 15 NUREG-2246 I think does a pretty good job of laying 16 out the landscape of how fuel qualification works in 17 terms of regulatory requirements. It provides a lot 18 of the technical basis for how you would show that you 19 meet the regulatory requirements.
20 But there arent necessarily a ton of 21 regulatory requirements that directly apply to fuel 22 qualification as a process. But 50.43E requires your 23 safety features to be supported by analysis testing 24 operating experience of a combination thereof. And it 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
99 requires there to be sufficient data to exist to 1
assess your analytical tools.
2 And then 5034 requires applicants to 3
evaluate a postulated of fission probabilities from 4
the core in the containment. Part of that is the 5
fuels performance. And as TerraPower mentioned, that 6
is something that they are doing with their major 7
accident as part of their construction permit 8
application.
9 And it requires the principal design 10 criteria to be submitted. Some of the PDCs have to 11 fuel, so.
12 So then the guidance, theres NUREG-2246, 13 which provides general guidance on fuel qualification 14 for non-light water reactors in the form of this fuel 15 qualification assessment framework. And Ill be kind 16 of stepping through that a little bit today.
17 And that kind of, that draws on a lot of 18 the experience from the staff evaluating both light 19 water and non-light water reactor fuels. And then 20 also we have this NUREG-CR 7305, which was developed 21 by staff from, I think it was INL, Los Alamos, and 22 ANL, giving us some insights into metallic fuel 23 systems.
24 And it did that in the NUREG-2246 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
100 framework and identified an operating and low-key 1
behaviors and phenomena and provided a review of the 2
data that was available and discussed a little bit the 3
current state of fuel-performed follow-up.
4 And some of the key conclusions from the 5
NUREG-CR, I wont go through the whole thing in detail 6
because I did that during the subcommittee, and it 7
took a solid 20 minutes. But for fuel with geometry 8
and operating conditions consistent with the previous 9
operating experience, so thats really EBR-II and the 10 FFTF, MFF fuel, the metallic fuel that was operated at 11 FFTF.
12 The life-limiting and safety-related fuel 13 behaviors and well known and predictable, up to around 14 10 percent burnup. And thats not really a hard 15 limit, thats a, you know, we think its well-16 characterized up to this limit. Somewhere beyond 17 that point, the behaviors are less predictable. And 18 so if you wanted to go much beyond that, you would 19 need to do a more thorough job of characterizing it 20 than has been done previously.
21 Fuel constituent redistribution is one of 22 the behaviors that is present in the data that does 23 affect fuel properties and other things. Thats 24 captured in the existing empirical models that 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
101 based on this fuel operating experience. The life-1 limiting phenomenon is fuel cladding chemical 2
interaction, which TerraPower mentioned as being their 3
main issue that theyre dealing with.
4 And fuel cladding mechanical interaction 5
is not really a concern. But again, thats really 6
specific to, you know, similar geometry to what the 7
previous operating experience was, and but the lower 8
end of the burnups that were operated.
9 Transient data would help to establish 10 safety margins. TerraPower talked about doing 11 additional transient testing. And that if you wanted 12 to use a highly mechanistic model, you would need to 13 do more work to qualify that. So for example, you 14 know, what effect does fuel constituent redistribution 15 have. Thats something that you would need to study 16 a little bit more closely.
17 So the Natrium fuel assembly design, its 18 very similar to the EBR tool and that MFF fuel from 19 FFTF. Its a U-10Zirc peak enrichment less than 20%,
20 so its HALEU. Seventy-five percent smear density.
21 These are all essentially the same characteristics 22 that are discussed in the NUREG-CR.
23 TerraPower showed the assembly overview.
24 You saw the hexagonal fuel assembly. And then theyre 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
102 applying this limited free bow core restraint system, 1
which was sort of tried at FFTF. Or I guess was, you 2
could say was tried at FFTF. So there is some 3
information that they can validate against there.
4 Now, Ill just start walking through the 5
fuel qualification big framework from NUREG-2246. So 6
just it has this top level goal that fuel is qualified 7
for use, and thats supported by all of these 8
different subgoals. So Goal 1.1 and 1.2, or really 9
all of Goal 1 is talking about the fuel manufacturing 10 and whether thats in an appropriately controlled and 11 understood process.
12 Our take on all of this was that the TR 13 either includes or refers to design documents that 14 TerraPower has that have this information to we think 15 an appropriate degree. They did mention in their 16 topical report that theres the potential for fuel --
17 or for materials other than U-10Zirc or HT-9 to be 18 part of the fuel system.
19 We included a limitation and condition on 20 there to essentially say if you are going to use these 21 materials, you need to describe them a little bit 22 more. But I will say that all the materials that they 23 mentioned in the topical report are, you know, code-24 qualified materials that are generally used in 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
103 industry in these kind of applications. So theyre 1
not things that were particularly concerned about.
2 One thing thats important for fuel is, 3
you know, making sure that the end-state attributes 4
from the manufacturing process are appropriately 5
captured. And we thought that TerraPower did that 6
well enough in the topical report.
7 So Goal 2 talks about margin to safety 8
limits, and thats supported by design limits for 9
normal operation of AOOs and then also for accidents.
10 Part of that is defining the fuel performance envelope 11 that you want to be working in. TerraPower provided 12 those in a pin and assembly damage criteria that they 13 flashed on the screen earlier, and that was consistent 14 with the key mechanisms that we saw from the NUREG.
15 We havent seen specific limits on any of 16 those criteria yet, so thats something that we would 17 need to better understand before the fuel is 18 considered to be fully qualified. And Ill talk a 19 little about how their operating envelope compares to 20 the historical operating experience later.
21 And one thing here I grayed out evaluation 22 model is available. They included a discussion on 23 evaluation models. So we know that they have 24 analytical methods to assess the fuel. We didnt 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
104 really review them in this topical report.
1 We think that the methods that they have 2
look like they have what they need, you know, in terms 3
of geometry and fields and stuff like that to be able 4
to model the fuel. But its not like this topical 5
report had a validation of those methods, because that 6
data is still being collected. So thats something 7
that were going to have to deal with later on.
8 So we already talked about 211 on the 9
previous slide. Then this talks about these, so these 10 are the release limits under accident conditions. And 11 I just wanted to mention here these two bullets that 12 I wanted to highlight really relate to limitation and 13 condition 5, and thats the specifying the retention 14 and release requirements.
15 TerraPower said that those -- that was 16 going to be done in the mechanistic source term 17 topical report. So it was outside the scope of the 18 fuel topical report review. And so we are actively 19 reviewing that topical report.
20 And here I will talk about the safety 21 limits for accidents and -- transients and accidents.
22 So the fuel failure criteria that TerraPower came up 23 with were we thought consistent with the key 24 mechanisms that we identified for metallic fuel. 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
105 coolable geometry criteria was also consistent.
1 There was a discussion in the NUREG-CR 2
about ejection of molten debris from the fuel. They 3
-- we thought that those were precluded by having a 4
limit against fuel melt. If youre not going to melt 5
the fuel, there isnt really a mechanism to eject much 6
debris from the fuel. Theres a lot of run beyond 7
cladding breach testing that shows that the fuel just 8
kind of sits there and nothing really happens to it.
9 The negative reactivity insertion criteria 10 we thought were adequate. But again, as with the 11 discussion on AOOs and normal operation, we would 12 still need to understand what the specific limits 13 would be on these criteria.
14 I already touched on the evaluation model, 15 so Ill just skip through this slide. Data, so there 16 is, as TerraPower mentioned, a lot of historical data 17 out there. We focused in our safety evaluation on the 18 scope and applicability of the previous data and how 19 that data supports TerraPowers acceptance criteria, 20 which I will say they did a really good job of laying 21 out in the topical report, you know, how --what 22 testing supports each criterion.
23 The type I fuel design and geometry that 24 they have is generally consistent with metallic fuel 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
106 that was operated at EBR-II and FFTF. Its a little 1
bit fatter than the EBR-II fuel, a little, just a tiny 2
bit fatter than the FFTF fuel. You saw the length of 3
the fuel columns is about the same as FFTF.
4 A significantly larger plenum, which is 5
good for accommodating fission gas release. So those 6
differences we think are either beneficial in terms of 7
the plenum. And I think the fuel cladding is slightly 8
thicker too. Or arent expected to have much impact 9
on the applicability of the historical data. Theyre 10 small deltas.
11 The fuel operating parameters were also 12 generally consistent with the past operating 13 experience. Some of those parameters are at or maybe 14 slightly beyond the historic database. But those 15 deltas we think are small, and theyre not expected to 16 have a lot of effect. They would be addressed by the 17 surveillance program or are covered by testing that 18 TerraPower proposed to do.
19 And our overall conclusion is that the new 20 data collection, so basically the historical data is 21 generally applicable. Where there are gaps, the new 22 data collection that TerraPower proposed is 23 appropriate to fill those gaps.
24 Shifting gears a little bit to talk about 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
107 the control assembly, so theyre boron carbide pellets 1
in a plenum that looks a little, or in a fuel --
2 control rod, not a fuel rod, that looks a little bit 3
more sort of like an LWR rod, where its got a plenum 4
with a hold-down spring.
5 But they also are clad in HT-9 with an HT-6 9 wire wrap, like the fuel rods are. And theyre in 7
that sort of tight, hexagonal arrangement.
8 The one thing that is important about the 9
control rod design that isnt necessarily super 10 obvious from the discussions that weve had already is 11 that each control assembly occupies its own space in 12 the core with its own duct. There is then a control 13 rod duct inside that duct that moves up and down.
14 And so they, as we talked about during the 15 previous meeting, you know, theres primary and 16 secondary control assemblies to try to meet that PDC-17 26 criterion. The differences are really the number 18 of absorber pins and the dimensions of the control 19 assembly.
20 And then I just have I think a single 21 slide on qualification of the control assemblies. But 22 you know, sort of boiling down the NUREG-2246 criteria 23 arent exactly applicable to a control assembly 24 because it has different safety functions. But you 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
108 kind of think of analogous criteria, at least at a 1
high level.
2 So you know, are there are appropriate 3
controls on manufacturing? Yeah, we looked at what 4
they provided in the topical report and what they 5
referenced in terms of design documents. We got to 6
look at those in an audit. We think that theyve 7
appropriately specified what the manufacturing looks 8
like to, at least to the degree that it needs to be.
9 The design criteria we thought were all 10 appropriate to make sure that the control rods could 11 fill -- fulfill their safety function.
12 For evaluation model, kind of similar 13 story as with the fuel rods where the codes we think 14 have the ability to do what they need to do, but 15 theres some validation that still needs to happen.
16 And essentially I think that, as was mentioned during 17 the TerraPowers meeting, its the same codes, but 18 they wanted -- they added boron carbide models.
19 For data, there is some historical data 20 from past operating fast reactors for different 21 control rod performance that TerraPower was able to 22 draw on. I would say theres no exact one-to-one 23 match for control rods in terms of like materials.
24 And but there are -- there are some that use different 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
109 combinations.
1 So we thought that the data that 2
TerraPower was able to assemble looked like it covered 3
the spectrum well enough. And they have planned 4
testing again to fill gaps in this historical 5
database.
6 Talked briefly about fuel surveillance and 7
LDAs and LTAs, TerraPower touched on this. Theres a 8
notional surveillance plan for the first several 9
cycles of irradiation in the topical report. The LDAs 10 and LTAs are designed with removable pins to 11 facilitate close irradiation examination.
12 Theres significant precedent for a 13 program like that, LDAs and LTAs, based on the 14 operating fleet. We do want to see eventually more 15 detail on how those leak demonstration, leak test 16 assemblies will be evaluated.
17 To the point that you brought up, you 18 know, the removable pins wont have wire wrap. So how 19 does that affect the performance of those pins and how 20 do you evaluate it? Thats something that wasnt 21 necessarily clear from the topical report. So thats 22 something were going to dig into as we go forward.
23 Limitations and conditions. So the first 24 one really is, you know, this is a good plan. But 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
110 that does -- its a plan and you still need to execute 1
it. The second one is the point that I touched on 2
about use of materials other than U-10Zirc and HT-9 in 3
fuel. If you are going to use those, we need to talk 4
more about it.
5 The topical report, sorry, the third 6
criterion here relates to the relationship between the 7
fuel design limits and the SRDLs. Essentially this is 8
good fuel design limits that provides good context for 9
evaluating the SRDLs. But you have to actually 10 evaluate the SRDLs for like stochastic failures of 11 fuel pins or whatever.
12 For number 4, and I can talk more about 13 this if we want to have a closed session, but I did 14 cover it during the subcommittee meeting. There were 15 some documents that TerraPower referred to in the 16 topical report for helping develop their design 17 criteria that hadnt yet been the subject of NRC 18 reviews. So we just wanted to point that, that 19 criterion there or that L&C there.
20 And limitation 5 really relates to the 21 retention of radionuclides. And we think that its 22 okay to push that off to a different topical report, 23 we just wanted to put this limitation to make it clear 24 what the scope of this topical is.
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111 And our overall conclusion is that the 1
topical report was acceptable and provided an overall 2
acceptable approach for qualifying fuel and control 3
assemblies.
4 And the one thing I will say, and this did 5
come up last time, part of that is the monitoring and 6
surveillance program we think is a really important 7
part of the overall fuel qualification effort.
8 CHAIR KIRCHNER: Go back to No. 5, please.
9 So you would expect the actual performance 10 would be in the mechanistic source term report?
11 MR. ANZALONE: So we left it open. I 12 dont think we said this has to be in the mechanistic 13 source term. But I think TerraPower has said that 14 its covered by the mechanistic source term topical 15 report.
16 CHAIR KIRCHNER: Just always a little bit 17 on guard, so to speak, when you have statements like 18 are expected to remain within the fuel, etc. So this 19 implies that theyre going to demonstrate that or make 20 the case somewhere else.
21 MR. ANZALONE: Yeah, in their evaluation 22 of the source term, they would have to justify 23 whatever is happening to the radionuclides. If 24 theyre crediting retention, say, like I think we have 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
112 a reasonable expectation that a lot of radionuclide, 1
especially solid fission products are going to be 2
retained within the fuel matrix. So thats just based 3
on, you know, the data thats out there.
4 But if TerraPower wants to credit that in 5
their mechanistic source term analysis, thats 6
something that theyre going to have to talk about at 7
that point. Thats what this limitation --
8 CHAIR KIRCHNER: Yeah.
9 MEMBER PETTI: It doesnt imply that there 10 isnt any release from cladding breach. Some fission 11 products are coming out into the sodium, sure.
12 MR. ANZALONE: Yeah.
13 MEMBER PETTI: This data. Yeah, but 14 theres a lot of other fission products --
15 MR. ANZALONE: Exactly.
16 CHAIR KIRCHNER: Steve, you have your hand 17 up.
18 DR. SCHULTZ: Yes, thank you. Reed, you 19 mentioned as you described the -- in particular the 20 methodologies that are being used to evaluate the 21 control rod performance, control element performance.
22 Do you -- can you expand on that, what 23 youre looking for in terms of what additional work 24 needs to be done there and when we can expect 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
113 information to come forward? Is it benchmarking, or 1
something more than that?
2 MR. ANZALONE: So thats a good question.
3 I think in general, like we -- Im comfortable with 4
the state of things as far as having like preliminary 5
analyses to support the PSAR. I think we would want 6
the sort of more full qualification to be done before 7
the operating license, by the operating license.
8 I dont know if that answers your 9
question, though.
10 DR. SCHULTZ: Well, you specifically 11 mentioned that the methodologies would need additional 12 attention. And is that what youre referring to 13 there, that --
14 MR. ANZALONE: Yeah, yeah, that we would 15
-- that we would need to have some way of validating, 16 right, that. So say, you know, you -- theres going 17 to be a pressurization of the control rods as you, you 18 know, burn up the boron, for lack of a better word.
19 So you would want to be able to make sure that those 20 arent going to break open and spill out all of their 21 poison.
22 So we would need to be able to see 23 eventually an evaluation of that and have some 24 confidence that the models were validated for that.
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114 MEMBER PETTI: Also I would think the 1
dimensional change --
2 MR. ANZALONE: Yep.
3 MEMBER PETTI: Because of this in the past 4
some either fuel or control assemblies stick and --
5 DR. SCHULTZ: Yes.
6 MR. ANZALONE: Yeah.
7 MEMBER PETTI: Theres those issues that 8
9 MR. ANZALONE: No, there are a lot --
10 there are a lot of different issues, yeah, absolutely.
11 I was just giving that as an example.
12 DR. SCHULTZ: Thank you, thats helpful.
13 Appreciate it.
14 MR. ANZALONE: But yeah, I think the big 15 one is like dimensional change. And you can think 16 about that either at like a pin level, right, you have 17 swelling of the pin, and then maybe that stops there 18 being appropriate cooling of the adjacent pins in the, 19 you know. They have essentially subchannels too, like 20 the fuel does. Or, at the assembly level you get 21 deformation that stops it from being able to insert.
22 So thats definitely something that we 23 want to pay attention to, because we think its --
24 thats the key thing that drives the control rod 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
115 insertability, is the assembly-level deformation.
1 DR. SCHULTZ: We talked in detail about 2
the fuel, the fuel assembly qualification, fuel rod 3
qualification program thats been proposed. Are you 4
comfortable with what has been proposed with respect 5
6 MR. ANZALONE: Yeah. And I just didnt 7
talk about it in as much detail in this presentation.
8 I would say because the criteria are different, its 9
-- theyre a little less tight because of the nature 10 of control rods and their design function. But I 11 would say that theres basically just as much in the 12 topical report about control assembly qualification as 13 there is fuel.
14 DR. SCHULTZ: Good, thank you.
15 MEMBER PETTI: So we, I know we talked 16 about this in subcommittee. The whole qualification 17 runs through all these codes. A heck of a lot of 18 computer codes need a lot of data validation. And 19 that always makes me a little bit nervous.
20 MR. ANZALONE: Yeah.
21 MEMBER PETTI: What Im hoping is that the 22 margin that, from an engineering gut feel that you 23 have when you look at these designs, you look at 24 performance, that that can translate through those 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
116 codes to give you the analytical margins that you need 1
when you have to go 95% confidence and stack up all 2
these (audio interference).
3 Thats the one thing that its hard to 4
see, the report doesnt really get into that at all.
5 But its the one thing that I worry about that when 6
you get -- you dont know until you get --
7 MR. ANZALONE: Well, so its on our minds 8
too. I will say that that is one of the things that 9
were focused on looking at. Not this specific -- I 10 mean, its a through line for this topical report, 11 right.
12 But its as were looking at like their 13 design basis accident analysis methodologies, you 14 know, were thinking about what are their criteria 15 that are in there for fuel failure and how are they 16 actually evaluating that. So youre going to see more 17 of that as we come through the reviews.
18 MEMBER PETTI: So you know, I went back 19 and read the SER on PRISM, and theres an appendix 20 that they did, the staff had some of our labs do 21 calculations. And frankly the results, were talking 22 1990s, really quite good comparing GE and lab tools.
23 Granted, on reactivity they used the same reactivity 24 coefficients, but still the results were really, 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
117 really quite good.
1 But it never got into the details of fuel 2
model.
3 MR. ANZALONE: Sure.
4 MEMBER PETTI: And, you know, theres 5
always such a mixture of empiricism and semi-6 empiricism. And now theres better models, but it 7
remains to be seen that the sharper pencil gets you 8
the answer you want.
9 This has always been one of my concerns 10 about these really cool advanced models. Hopefully 11 they verify your engineering judgment. But that all 12 of that effort gets you margin and all of that in the 13 end you stack it all together.
14 MR. ANZALONE: Yeah, totally agree. Im 15 100% aligned on that.
16 MEMBER PETTI: Good.
17 MR. ANZALONE: And you know, one thing 18 Ill say is that were talking to the Office of 19 Research about ways in which they can support us with 20 doing confirmatory analyses and stuff like that, as 21 was done for the PRISM review.
22 Some of those I think would use our codes, 23 some of those might use, depending on where things go, 24 you know, the NEAMS codes like BISON or what have 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
118 for fuel performance, so.
1 MEMBER PETTI:
Theres some good 2
publications already out there on BISON, the amount of 3
fuel, that I found really helpful.
4 MR. ANZALONE: And I think one thing that 5
I am trying to be cognizant of when we have those 6
conversations is, you know, to what degree are --
7 because this is something that was brought up in that 8
NUREG-CR, to what degree are those mechanistic models 9
actually well-validated and is there the data to 10 support them. I think that it kind of remains to be 11 seen a little bit.
12 But TerraPower, I think that their 13 approach that theyre taking is solid, so, not too 14 concerned with their modeling approach here.
15 MEMBER ROBERTS: If theres no more 16 questions from members or consultants? I guess its 17 time now to go out for public comments.
18 If theres any members of the public whod 19 like to make a comment, please go ahead and unmute 20 yourself, state your name and affiliation if there is 21 one, and then state your comment, please.
22 Hearing none, guess Ill turn the meeting 23 back over to Chair Kirchner.
24 CHAIR KIRCHNER: Thank you, Tom and Dave.
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
119 Thank you to the presenters, both staff and Applicant, 1
thank you.
2 And at this point, were actually a little 3
ahead of schedule for quite a change. And so weve 4
set aside a period now to have committee deliberation 5
on what we heard on both of these topical reports.
6 And then we can proceed at this point with our letter 7
writing.
8 So, Jose, would you like to make a 9
comment?
10 MEMBER MARCH-LEUBA: Court reporter, is he 11 needed the rest of the week?
12 CHAIR KIRCHNER: Let me confer with Larry.
13 Do, at this point do we need the court reporter 14 further?
15 MR. BURKHART: I think were going into 16 deliberation and letter writing. We can let the court 17 reporter loose.
18 CHAIR KIRCHNER: Looking at the schedule 19 for today and tomorrow, and --
20 MR. BURKHART: It is all we have left, 21 yes.
22 CHAIR KIRCHNER: We P&P tomorrow, and so 23 we --
24 MR. BURKHART: Correct.
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
120 CHAIR KIRCHNER: Dont normally record 1
that, correct?
2 MR. BURKHART: We dont, no.
3 CHAIR KIRCHNER: With that, okay.
4 For the court reporter, thank you. I 5
dont believe that well need your services for the 6
rest of this meeting and this week.
7 (Whereupon, the above-entitled matter went 8
off the record at 11:11 a.m.)
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
Staff Review of NATD-FQL-PLAN-0004, Fuel and Control Assembly Qualification Reed Anzalone, Senior Nuclear Engineer Mallecia Sutton, Senior Project Manager Office of Nuclear Reactor Regulation Division of Advanced Reactors and Non-Power Production and Utilization Facilities
Agenda
- Topical report (TR) purpose and review strategy
- Regulatory requirements and guidance
- Overview of NUREG/CR-7305
- Safety evaluation (SE) overview
- Fuel assembly design and qualification
- Control assembly design and qualification
- Limitations and conditions
- Conclusions 2
TR Purpose
- Provides plan to qualify Natrium Type 1 fuel (uranium-zirconium alloy in HT9 cladding) and Natrium control assemblies
- Requests NRC review and approval of the following:
- Acceptance criteria are adequate to support fuel qualification
- Identified key manufacturing parameters are adequate to support fuel qualification
- Evaluation methods and models are adequate to support fuel qualification
- Use of legacy data and planned testing are adequate to provide necessary information for qualification of the fuel
- Planned use of pins outside the performance envelope of the bulk of the core or that advanced design features are acceptable
- Presents select fuel qualification results and ongoing and planned qualification activities 3
TR Review Strategy
- Review scope and adequacy of fuel qualification plan in the context of NUREG-2246, Fuel Qualification for Advanced Reactors (ML22063A131)
- Review technical details of fuel and qualification efforts against information in NUREG/CR-7305, Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, Fuel Qualification for Advanced Reactors (ML23214A065) 4
Regulatory Requirements
- Requires safety features to be supported by analysis, testing, operating experience, or a combination thereof.
- Requires sufficient data exists to assess analytical tools
- Requires applicants to evaluate a postulated fission product release from the core into containment
- Requires principal design criteria (PDCs) to be submitted NUREG-2246: Fuel qualification provides a means to identify safety criteria for the fuel, which then are used to establish performance criteria for facility structures, systems, and components (SSCs). Facility safety is then addressed by description and analyses of these SSCs.
5
Guidance
- NUREG-2246, Fuel Qualification for Advanced Reactors
- Provides general guidance on fuel qualification for non-light water reactors (non-LWRs) in the form of a Fuel Qualification Assessment Framework (FQAF)
- NUREG/CR-7305 Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, Fuel Qualification for Advanced Reactors
- Provides a generic response to NUREG-2246 for a uranium-zirconium metal fuel system, including
- Identification of an operating envelope and key behaviors/phenomena
- Review of available data
- Discussion of current state of fuel performance modeling 6
NUREG/CR-7305 - Key Conclusions
- For fuel with geometry and operating conditions consistent with previous operating experience, life-limiting and safety-related fuel behaviors are well known and predictable up to 10 atom-% burnup
- Fuel constituent redistribution is captured in data
- FCCI is life-limiting phenomenon
- FCMI is not a concern
- Additional transient data would help to establish safety margins
- More work is needed to qualify mechanistic models 7
Natrium Fuel Assembly Design
- Very similar to EBR-II fuel and metallic fuel operated at the Fast Flux Test Facility (FFTF) 8
- Pin characteristics:
- Peak enrichment < 20%
- 75% smear density
- Sodium bond
- HT9 cladding
- Axial shield slug
- Large plenum
- HT9 wire wrap
- Assembly characteristics
- Pins arranged in tight triangular pitch in hexagonal bundle
- Hexagonal duct, inlet nozzle, handling socket
- Limited free bow core restraint system
NUREG-2246 FQAF
- G1. Fuel is manufactured in accordance with a specification
- G1.1 Key dimensions and tolerances of fuel components are specified
- G1.2 Key constituents are specified with allowance for impurities
- TR includes/refers to adequate design information
- L&C #2 covers use of materials other than U-10Zr/HT9
- G1.3 End state attributes for materials within fuel components are specified or otherwise justified.
- Adequate end state attributes provided 9
NUREG-2246 FQAF
- G2. Margin to safety limits can be demonstrated.
- G2.1 Margin to design limits can be demonstrated under conditions of normal operation and AOOs.
- G2.1.1 Fuel performance envelope is defined
- Pin and assembly damage criteria consistent with key mechanisms from NUREG/CR-7305
- Specific limits must be provided before fuel is considered qualified (L&C #1)
- Comparison to fuel operating experience discussed later
- G2.1.2 Evaluation model is available 10
NUREG-2246 FQAF
- G2. Margin to safety limits can be demonstrated.
- G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated.
- G2.1.1 Fuel performance envelope is defined
- G2.2.1 Radionuclide retention requirements are specified
- Addressed in separate TR; L&C #5
- G2.2.2 Criteria for barrier degradation and failure are suitably conservative
- G2.2.3 Radionuclide retention and release from fuel matrix are modeled conservatively
- Addressed in separate TR; L&C #5 11
NUREG-2246 FQAF
- G2. Margin to safety limits can be demonstrated.
- G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated.
- G2.2.2 Criteria for barrier degradation and failure are suitably conservative
- G2.3 Ability to achieve and maintain safe shutdown is assured.
- G2.3.1 Coolable geometry is ensured
- G2.3.2 Negative reactivity insertion can be demonstrated
- Fuel failure criteria consistent with key mechanisms from NUREG/CR-7305
- Coolable geometry criteria are consistent with NUREG/CR-7305, except molten debris ejection which is precluded by preventing fuel melt
- Negative reactivity insertion criteria are adequate
- Specific limits must be provided before fuel is considered qualified (L&C #1) 12
Evaluation Models
- Separate EM assessment framework in NUREG-2246
- TR does not contain detailed information on fuel performance models, staff did not fully assess against NUREG-2246 framework
- Codes discussed in TR appear to provide the capabilities needed to support fuel qualification efforts
- Additional effort is needed to demonstrate that the proposed EMs contain all necessary material and physics models, verify the EMs, and validate them against experimental data.
- EMs will be evaluated in future revision of this TR or in a separate TR specifically covering fuel performance 13
Data
- Because evaluation of historical data and data collection is ongoing, focus in SE is on scope and applicability of historical data, how data supports TerraPowers acceptance criteria, and plans for future testing
- Type 1 fuel design geometry is generally consistent with metallic fuel operated at EBR-II and FFTF
- Differences either beneficial or not expected to have much impact on applicability of historical data
- Fuel operating parameters also generally consistent with EBR-II/FFTF
- Some parameters are at or slightly beyond historical database
- Deltas are small, and are not expected to have much effect, will be addressed by TerraPowers planned surveillance program, and/or will be the subject of proposed testing discussed in the TR
- New data collection appropriate to fill gaps 14
Control Assembly Designs
- Pin characteristics
- Natural boron carbide pellets
- Plenum with spring
- HT9 cladding
- HT9 wire wrap
- Assembly characteristics
- Triangular pitch in hexagonal lattice
- Upper guide plate with coupling head
- Control rod duct that moves up and down inside control assembly duct
- Primary/secondary control assembly differences
- Number of absorber pins
- Dimensions, including space between inner control rod duct and control assembly duct 15
Control Assembly Qualification
- Manufacturing
- Control assembly manufacturing appropriately specified
- Design criteria
- Damage, failure, and insertability criteria adequate to ensure control rods can fulfil their safety function
- Evaluation model
- Same codes as fuel assemblies with changes for control assemblies
- Codes appear capable but more work is needed
- Data
- Historical data from EBR-II, FFTF, Joyo
- Planned testing to fill gaps in historical data 16
Fuel Surveillance, LDAs, and LTAs
- TR presents notional surveillance plan for first several cycles
- LDAs and LTAs designed with removable pins to facilitate post-irradiation examination (PIE)
- Significant precedent for LDA/LTA program based on operating fleet
- Additional detail required on how LDAs/LTAs will be evaluated and how uncertainties in performance will be captured in analyses 17
L&Cs
- 1. This TR represents an acceptable approach for qualifying Natrium Type 1 fuel and control assemblies for use in a reactor but does not in and of itself demonstrate that the fuel and control assemblies are qualified. Additional activities, including those discussed in the NRC staffs SE, must be completed to execute this plan and appropriately justify that the fuel and control assemblies are qualified.
- 2. This TR addresses the material properties and performance of U-10Zr and HT9 in fuel. If other materials are used in the fuel system in licensing applications, the applicant or licensee must demonstrate that they are manufactured according to standard specifications and used consistent with their qualification under relevant NRC-accepted codes and standards, or otherwise appropriately justified.
18
L&Cs
- 3. This TR does not provide a means for demonstrating that proposed SARRDLs are satisfied during normal operations and AOOs for the Natrium plant. The role of the fuel acceptance criteria is to demonstrate that the fuel system is not damaged as a result of normal operations and AOOs; if these criteria are satisfied, then the fuel system need not be further assessed against the SARRDLs. However, the SARRDLs must still be evaluated against other sources of radionuclides, including circulating radionuclides resulting from an appropriate number of random fuel failures.
- 4. The (( )) have not been subject to previous NRC review or approval. If they are to be used to develop design criteria and associated limits that support fuel assembly acceptance criteria, these design criteria and associated limits must be appropriately justified.
19
L&Cs
- 5. This TR does not address the extent to which the fuel system is expected to retain radionuclides following a cladding breach. If an applicant or licensee wishes to qualify Natrium Type 1 fuel with an expectation that radionuclides are expected to remain within the fuel following a cladding breach, models for fuel system radionuclide retention and release must be proposed and appropriately justified by comparison to experimental data.
20
Conclusions TR is acceptable for referencing in future licensing submittals, subject to limitations and conditions.
- The NRC staff determined that the TR provides an acceptable approach for qualifying fuel and control assemblies for the Natrium reactor based on
- (1) the inclusion of sufficient information to demonstrate that fuel and control assemblies are manufactured in a process that provides adequate control over key parameters,
- (2) the identification of appropriate safety criteria for both fuel and control assemblies,
- (3) the development and justification of a significant applicable historical test database,
- (4) the development of a test plan that appropriately fills gaps in the historical test database, and
- (5) a robust fuel monitoring program, subject to the limitations and conditions discussed above.
Accordingly, the NRC staff concludes that the qualification plan provided in the TR can be used to support compliance with 10 CFR 50.43(e) and proposed Natrium PDCs.
21
Abbreviations ACCI - Absorber-cladding chemical interaction AOO - Anticipated operational occurrence CFR - Code of Federal Regulations EBR-II - Experimental Breeder Reactor-II EM - Evaluation model FCCI - Fuel-cladding chemical interaction FCMI - Fuel-cladding mechanical interaction FFTF - Fast flux test facility FQAF - Fuel qualification assessment framework LDA - Lead demonstration assembly LTA - Lead test assembly Non-LWR - Non-Light Water Reactor PDC - Principal design criterion PIE - Post-irradiation examination Pu - Plutonium RAC - Regulatory Acceptance Criteria SARRDL - Specified acceptable radionuclide release design limit SE - Safety evaluation SSC - Structure, system, or component TR - Topical report TRISO - Tri-structural Isotropic U - Uranium Zr - Zirconium 22
Review Chronology
- January 25, 2023: Submittal of TR Fuel and Control Assembly Qualification Plan, Revision 0 (ML23025A409)
- March 21, 2023: Pre-Application Public Meeting (ML23157A332)
- March 31, 2023: TR accepted for review by the NRC staff (ML23086C087)
- April 18, 2023: Submittal of correction to TerraPower Fuel and Control Assembly Qualification Topical Report (ML23109A099)
- June, July, and August 2023: Audit Conducted (ML24043A155)
- March 20, 2024: Draft SE Issued (ML24079A118) 23
NUREG-2246 FQAF 24 G. Fuel is qualified for use.
G1. Fuel is manufactured in accordance with a specification.
G1.1 Key dimensions and tolerances of fuel components are specified.
G1.2 Key constituents are specified with allowance for impurities.
G1.3 End state attributes for materials within fuel components are specified or otherwise justified.
G2. Margin to safety limits can be demonstrated.
G2.1 Margin to design limits can be demonstrated under conditions of normal operation and AOOs.
G2.1.1 Fuel performance envelope is defined G2.1.2 Evaluation model is available G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated.
G2.1.1 Fuel performance envelope is defined G2.2.1 Radionuclide retention requirements are specified G2.2.2 Criteria for barrier degradation and failure are suitably conservative G2.2.3 Radionuclide retention and release from fuel matrix are modeled conservatively G2.3 Ability to achieve and maintain safe shutdown is assured.
G2.3.1 Coolable geometry is ensured G2.3.2 Negative reactivity insertion can be demonstrated
NUREG/CR-7305 Design Parameters
- Uranium-10 weight% zirconium alloy fuel
- 75% smear density
- 1.4 plenum to fuel volume ratio
- Sodium bond
- HT9 cladding
- Fuel dimensions from Experimental Breeder Reactor-II (EBR-II) 25
NUREG/CR-7305 - Fuel Geometric Evolution 26
- Fuel swells axially and radially until cladding contact
- Porosity interconnects and gaseous fission products are released to the plenum
- Solid fission product build up
- At >10 atom% burnup, fission gas flow through pores becomes constrained and fuel begins to swell again
NUREG/CR-7305 - Fuel Constituent Redistribution
- Thermal gradient in fuel drives redistribution of U and Zr in fuel
- Higher operating temperatures and linear heat rates drive more redistribution
- Potentially affects fuel properties, local power density
- Accounted for in experimental data below 10% burnup 27 Kim, Yeon Soo, S. L. Hayes, G. L. Hofman, and A. M. Yacout. "Modeling of constituent redistribution in U-Pu-Zr metallic fuel." Journal of Nuclear Materials 359, no. 1-2 (2006): 17-28.
NUREG/CR-7305 - Cladding Integrity/Barrier Degradation
- Fuel-cladding mechanical interaction (FCMI) not a concern below 10% burnup for fuels with 75% smear density
- Fission gas release not a concern with appropriately sized plena
- Fuel-cladding chemical interaction (FCCI) is primary source of cladding degradation and fuel failure
- Thins cladding due to formation of low-melting point eutectics at fuel-cladding interface
- U-Fe but also contributed to by lanthanides, which tend to migrate down thermal gradient
- Measurable thinning at ~725°C, NUREG/CR recommends steady-state limit of 650°C 28
NUREG/CR-7305 - Fuel Properties
- Porosity and redistribution evolution affect properties
- Significant margin to solidus temperature (>1100°C); bulk fuel melting is not a concern and FCCI region provides limit for fuel temperature
- Limited thermal conductivity data but favorable compared to UO2
- Limited irradiated mechanical properties but below 10% burnup, empirical models adequately predict fuel swelling and cladding strains 29
- Transient testing (in-pile and out-of-pile) has been done and identified FCCI to be the primary failure mode
- Additional transient testing is needed to characterize operating envelope 30
G1.1 & G1.2 - Key Dimensions & Constituents
- TR refers to design drawings and materials specifications
- TR also includes details on HT9 and U-10Zr composition
- Staff audited referenced documents and found that they contained appropriate information.
- Use of materials other than U-10Zr and HT9 not clear - Limitation and Condition (L&C) 2 31
G1.3 - End-State Attributes
- NUREG/CR-7305 provides details on manufacturing process and important end-state attributes for U-10Zr, summarized as:
- Injection molding with controls on formation of oxides and fuel density
- Fuel rod plenum sized appropriately
- Manufacturing process discussed at high level in TR, with references to specifications, including fabrication process
- Consistent with end-state attributes discussed in NUREG/CR 32
G2.1.1 - Fuel Performance Envelope
- TerraPower developed Regulatory Acceptance Criteria (RAC) for different mechanisms to provide an envelope in which fuel damage can be precluded
- Damage: Fuel has not failed but may have reduction in functional capability (i.e., outside of safety analysis assumptions) 33
G2.1.1 - Fuel Performance Envelope
- Pin Damage Criteria
- Stress, strain, loading
- Fatigue
- Fretting wear
- Erosion and corrosion
- Cladding damage due to FCCI
- Dimensional changes (rod bowing or swelling)
- Pin internal pressure
- Fuel and cladding temperatures
- Assembly Damage Criteria
- Stress, strain, loading
- Fatigue
- Fretting wear
- Erosion and corrosion
- Dimensional changes (duct bowing and dilation)
- Hydraulic loads exceeding hold-down
- Assembly component temperatures 34
G2.1.1 - Fuel Performance Envelope
- Damage mechanisms presented are consistent with key phenomena and properties from NUREG/CR-7305
- Staff did not evaluate specific limits to prevent damage, which are expected to be under development as part of fuel qualification plan (L&C 1)
- Operating envelope and comparison to historical data is discussed in more detail in experimental data assessment framework 35
G2.1.1 - Fuel Performance Envelope (Accidents)
- TerraPower developed separate RAC for accidents; these are assessed under separate goals for barrier failure, radionuclide retention and release, coolable geometry, and negative reactivity insertion 36
- Addressed in separate TR (source term methodology)
- L&C 5 G2.2.1 - Radionuclide retention requirements G2.2.3 - Radionuclide release modeling
G2.2.2 - Barrier Degradation & Failure Criteria
- Barrier degradation criteria covered under G2.1.1
- Pin failure criteria include:
- Cladding and slug overheating
- For gross melting but also rapid eutectic penetration
- Cladding deformation due to mechanical loads
- Fuel system mechanical fracturing from externally applied forces
- Cladding wastage (including wear, erosion, corrosion, FCCI, eutectics)
- Consistent with discussion in NUREG/CR-7305
- Future work to establish appropriate limits (L&C 1)
- Supporting data discussed in separate framework 37
G2.3.1 - Coolable Geometry
- TerraPower developed separate RAC related to coolable geometry:
- Stress and strain limits to ensure coolability
- Cladding and fuel temperatures below melting point
- Coolability evaluations must include cladding ballooning
- Structural deformation of fuel assemblies cannot prevent core cooling
- Hydraulic loads cannot unseat assemblies such that flow is reduced enough to prevent assembly cooling
- Generally consistent with NUREG/CR-7305, except debris ejected from failed fuel assemblies not explicitly addressed
- Based on historical data, preventing fuel melt precludes this issue 38
G2.3.2 - Negative Reactivity Insertion
- Negative reactivity insertion sensitive to control assembly distortion, unseating of control assemblies
- TerraPower developed separate RAC related to reactivity insertion:
- Structural deformation of control assemblies will not prevent the ability to insert control rods during accidents
- Hydraulic loads will not unseat control assemblies in a way that prevent insertion during accidents
- Other RAC also help ensure insertability:
- Fuel and control assembly distortion
- Fuel and absorber pin internal pressure
- Hydraulic loading on control assemblies
- Mechanical/neutronic design of control assemblies
- Criteria address possible mechanisms 39
NRC Staff Review of the Topical Report Principal Design Criteria for the Natrium Advanced Reactor, Revision 1 Stephanie Devlin-Gill, Senior Project Manager Reed Anzalone, Senior Nuclear Engineer Office of Nuclear Reactor Regulation Division of Advanced Reactors and Non-Power Production and Utilization Facilities
Agenda
- Topical Report (TR) purpose and review strategy
- Regulatory requirements
- Natrium principal design criteria (PDC) overview
- Key topics from ACRS subcommittee meeting
- Functional containment
- Specified acceptable system radionuclide release design limits (SARRDLs)
- Limitations and conditions 2
TR Purpose and Review Strategy
- Purpose of TR:
- Describe TerraPowers process for developing PDCs
- Provide PDCs to address compliance with Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a)(3)(i) for Construction Permit (CP) applications
- Describe rationale for meeting the intent of Natrium PDC 26, Reactivity Control Systems
- Review strategy
- Review Natrium PDC conformance with Regulatory Guide (RG) 1.232; group and evaluate deviations, considering key design features
- Identify interaction between RG 1.232 approach and Licensing Modernization Project (LMP)
- Review PDC 26 rationale 3
Regulations
- 10 CFR 50.34(a)(3)(i) requires an applicant for a CP to include the PDCs for the facility in the preliminary safety evaluation report (PSAR)
- 10 CFR 50, Appendix A provides requirements on the scope and content of PDCs for non-light water reactors (non-LWRs):
- The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
- These General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The General Design Criteria are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units.
4
Natrium PDC Overview
- TerraPower developed PDCs based on RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors (ML17325A611)
- 1-12, 14, 15, 17-19, 21-24, 26, 28-37, 44-46, 60-64, and 70-79
- 13, 16, 20, 25, 80, 81, and 82
- Used to implement functional containment or reflect use of SARRDLs
- No DC for 38-43, 50-57 due to use of functional containment 5
General Changes to PDCs A. Use of the term safety-significant B. Use of graded approach to coolant boundary quality C. Use of specified acceptable system radionuclide release design limit D. Use of functional containment concept E. Minor generic changes 6
A. Use of the term safety-significant B. Use of graded approach to coolant boundary quality C. Use of specified acceptable system radionuclide release design limit D. Use of functional containment concept E. Minor generic changes
Functional Containment Overview (1)
- RG 1.232, Appendix C, MHTGR-DC 16 (ML17325A611):
- The term functional containment is applicable to advanced non-LWRs without a pressure retaining containment structure. A functional containment can be defined as a barrier, or set of barriers taken together, that effectively limit the physical transport and release of radionuclides to the environment across a full range of normal operating conditions, AOOs, and accident conditions.
7
Functional Containment Overview (2)
- SECY-18-0096 (ML18115A157) documents approach to determining functional containment performance criteria
- Technology-inclusive, risk-informed, performance-based
- Methodology later developed into LMP
- Developed in parallel with RG 1.232
- SRM-SECY-18-0096 (ML18338A502) documents the Commissions approval of the NRC staffs approach to determining functional containment performance criteria for non-LWRs.
8
Non-Applicability of Containment Criteria
- TerraPower did not adopt DC 38, 39, 40, 41, 42, 43, 50, 51, 52, 53, 54, 55, 56, 57
- MHTGR-DC rationales note that these criteria are not applicable because there is not a pressure containing reactor containment structure
- Some relevant SFR-DC note that they would not be applicable if alternate approaches to containment were taken:
- SFR-DC 38: as necessary is meant to condition an SFR-DC 38 application to designs requiring heat removal for conventional containments that are found to require heat removal measures.
- SFR-DC 39 and 40 directly support 38
- SFR-DC 50 references a containment structure; 51-57 support 50 and state they are applicable to designs employing containment structures.
9
Natrium Functional Containment Considerations
- SECY-18-0096 and associated SRM indicates that functional containment concept is acceptable for non-LWRs
- Staffs finding is that certain reactor attributes are necessary for functional containment approach to be viable for Natrium; actual functional containment performance remains to be demonstrated
- Use of LMP implies method used to demonstrate functional containment performance:
- PRA and mechanistic source term analyses will be performed and must meet criteria (discussed in NEI 18-04, consistent with SECY-18-0096)
- Analyses will explicitly consider uncertainties
- Plant design will be evaluated for defense-in-depth adequacy per NEI 18-04 10
SARRDLs
- SARRDLs are for normal operation and AOOs, are established so 10 CFR Part 20 limits are not exceeded
- SECY-18-0096, Enclosure 2 (ML18115A367):
- Defining SARRDLs for specific designs is intertwined with functional containment performance criteria and would be developed by reactor designers as part of the integrated approach described in this enclosure.
11
Natrium SARRDL Considerations
- Staffs view is that SARRDLs are appropriate to use with functional containment and are consistent with a performance-based evaluation of releases
- TerraPowers fuel includes fuel design limits that can be used to help evaluate compliance with SARRDLs
- SARRDLs can be a useful tool for looking at ex-vessel events
- Means of monitoring would need to be included as part of design
- TerraPower must still propose and evaluate SARRDLs
- SARRDLs were discussed with TerraPower in a July 11, 2023, public meeting.
Closed discussion included examples.
12
Scope and Applicability of PDCs
- Proposed PDCs are based on RG 1.232 (traditional framework) but applied to licensing under NEI 18-04 (risk-informed, performance-based framework)
proposed PDC will need to address the functions provided by both SR and NSRST [non-safety related with special treatment] SSCs
- Proposed limitation 2 addresses potential gaps
- Will be addressed in CP application 13
Proposed Limitations and Conditions (L&Cs)
The NRC staff imposes the following L&Cs regarding the TR:
- 1. An applicant or licensee referencing this TR must propose a design that is substantially similar to the Natrium design as discussed in SE Section 1, or otherwise justify that any departures from these design features do not affect the conclusions of the TR and this SE.
- 2. The use of this TR is restricted to those applicants using the risk-informed, performance-based licensing process described in NEI 18-04, Revision 1, as endorsed by RG 1.233. Because the proposed PDCs may not fully address all performance requirements for SSCs defined as safety-significant under the NEI 18-04 process, applicants or licensees referencing this TR must augment the PDC in the TR with appropriate PDC for any SR or NSRST SSCs whose safety function relates to BDBEs, or NSRST SSCs needed for DID adequacy, or otherwise justify that the Natrium PDCs as described in the subject TR are adequate.
14
Conclusions
- TerraPower considered each of the design aspects presented in RG 1.232.
- TerraPower provided a sufficient set of PDCs for the Natrium design, subject to the L&Cs.
- The PDCs (subject to the L&Cs) establish the necessary design, fabrication, construction, testing, and performance DC for safety significant SSCs to provide reasonable assurance that the Natrium reactor could be operated without undue risk to the health and safety of the public.
- The TR is suitable for referencing in future licensing applications for the Natrium advanced reactor.
15
Abbreviations ARDC - Advanced reactor design criteria AOO - Anticipated operational occurrence BDBE - Beyond design basis event CFR - Code of Federal Regulations CP - Construction permit DANU - Division of Advanced Reactors and Non-Power Production and Utilization Facilities DC - Design criterion DBA - Design basis accident DBE - Design basis event GDC - General design criterion L&C - Limitation and/or condition LWR - Light water reactor MHTGR - Modular high temperature gas reactor NEI - Nuclear Energy Institute NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety related with special treatment NST - Non-safety related with no special treatment PDC - Principal design criterion PSAR - Preliminary safety evaluation report QA - Quality assurance RAC - Reactor air cooling system RG - Regulatory guide SAFDL - Specified acceptable fuel design limit SARRDL - Specified acceptable system radionuclide release design limit SFR - Sodium fast reactor SSC - Structure, system, or component SE - Safety evaluation SR - Safety related TR - Topical report 16
Natrium Functional Containment 17
Natrium SSCs Associated with Functional Containment Strategy
- Metallic fuel matrix and cladding
- Reactor enclosure system, head access area, and primary coolant boundary
- Sodium processing system
- Sodium cover gas system
- Intermediate heat transport system
- Reactor building
- Reactor auxiliary building
- Water pool fuel handling system
- Ex-vessel fuel handling system
- In-vessel fuel handling system
- Nuclear island heating, ventilation, and air conditioning system
- Gaseous radwaste processing system 18 Source: Kemmerer Unit 1 PSAR (ML24088A065)
TerraPower Approach to PDC Development 19 Topical Report Figure 1, PDC Development Flowchart
Natrium PDC - I. Overall Requirements Criterion Title Basis DC Modified?
1 Quality standards and records.
SFR-DC 1 Y - safety-significant 2
Design bases for protection against natural phenomena.
SFR-DC 2 Y - safety-significant 3
Fire protection.
SFR-DC 3 Y - safety-significant 4
Environmental and dynamic effects design bases.
SFR-DC 4 Y - safety-significant 5
Sharing of structures, systems, and components SFR-DC 5 Y - safety-significant, safe shutdown 20
Natrium PDC - II. Multiple Barriers Criterion Title Basis PDC Modified?
10 Reactor design.
SFR-DC 10 Y - SARRDLs 11 Reactor inherent protection.
SFR-DC 11 N
12 Suppression of reactor power oscillations.
SFR-DC 12 Y - SARRDLs 13 Instrumentation and control.
MHTGR-DC 13 Y - coolant boundary 14 Primary coolant boundary.
SFR-DC 14 Y - coolant boundary 15 Primary coolant system design.
SFR-DC 15 Y - coolant boundary 16 Containment design.
MHTGR-DC 16 Y - safety-significant 17 Electric power systems.
SFR-DC 17 Y - safety-significant, SARRDLs 18 Inspection and testing of electric power systems.
SFR-DC 18 Y - safety-significant 19 Control room.
SFR-DC 19 Y - safe shutdown 21
Natrium PDC - III. Reactivity Control 22 Criterion Title Basis PDC Modified?
20 Protection system functions MHTGR-DC 20 Y - safety-significant 21 Protection system testability and reliability.
SFR-DC 21 N
22 Protection system independence.
SFR-DC 22 N
23 Protection system failure modes.
SFR-DC 23 N
24 Separation of protection and control systems.
SFR-DC 24 N
25 Protection system requirements for reactivity control malfunctions.
MHTGR-DC 25 N
26 Reactivity control systems.
SFR-DC 26 Y - SARRDLs 27
[None - incorporated into 26 consistent with RG 1.232]
N/A N/A 28 Reactivity limits.
SFR-DC 28 Y - coolant boundary 29 Protection against anticipated operational occurrences.
SFR-DC 29 N
Natrium PDC - IV. Fluid Systems (1) 23 Criterion Title Basis PDC Modified?
30 Quality of primary coolant boundary.
SFR-DC 30 Y - coolant boundary 31 Fracture prevention of primary coolant boundary.
SFR-DC 31 Y - coolant boundary 32 Inspection of primary coolant boundary SFR-DC 32 Y - coolant boundary 33 Primary coolant inventory maintenance.
SFR-DC 33 Y - SARRDLs 34 Residual heat removal.
SFR-DC 34 Y - SARRDLs 35 Emergency core cooling.
SFR-DC 25 N
36 Inspection of emergency core cooling system.
SFR-DC 36 N
37 Testing of emergency core cooling system.
SFR-DC 37 Y - leaktight
Natrium PDC - IV. Fluid Systems (2) 24 Criterion Title Basis PDC Modified?
38
[Not used - functional containment]
N/A N/A 39
[Not used - functional containment]
N/A N/A 40
[Not used - functional containment]
N/A N/A 41
[Not used - functional containment]
N/A N/A 42
[Not used - functional containment]
N/A N/A 43
[Not used - functional containment]
N/A N/A 44 Structural and equipment cooling.
SFR-DC 44 Y - safety-significant 45 Inspection of structural and equipment cooling systems.
SFR-DC 45 N
46 Testing of structural and equipment cooling systems.
SFR-DC 46 Y - leaktight
Natrium PDC - V. Reactor Containment 25 Criterion Title Basis PDC Modified?
50
[Not used - functional containment]
N/A N/A 51
[Not used - functional containment]
N/A N/A 52
[Not used - functional containment]
N/A N/A 53
[Not used - functional containment]
N/A N/A 54
[Not used - functional containment]
N/A N/A 55
[Not used - functional containment]
N/A N/A 56
[Not used - functional containment]
N/A N/A 57
[Not used - functional containment]
N/A N/A
Natrium PDC - VI. Fuel and Reactivity Control 26 Criterion Title Basis PDC Modified?
60 Control of releases of radioactive materials to the environment.
SFR-DC 60 N
61 Fuel storage and handling and radioactivity control.
SFR-DC 61 Y - safety-significant 62 Prevention of criticality in fuel storage and handling.
SFR-DC 62 N
63 Monitoring fuel and waste storage.
SFR-DC 63 N
64 Monitoring radioactivity releases.
SFR-DC 64 Y - functional containment
Natrium PDC - VII. Additional PDC 27 Criterion Title Basis PDC Modified?
70 Intermediate coolant system.
SFR-DC 70 N
71 Primary coolant and cover gas purity control.
SFR-DC 71 N
72 Sodium heating systems.
SFR-DC 72 Y - safety-significant 73 Sodium leakage detection and reaction prevention and mitigation.
SFR-DC 73 Y - safety-significant 74 Sodium/water reaction prevention/mitigation.
SFR-DC 74 N
75 Quality of the intermediate coolant boundary.
SFR-DC 75 Y - safety-significant 76 Fracture prevention of the intermediate coolant boundary.
SFR-DC 76 Y - coolant boundary 77 Inspection of the intermediate coolant boundary.
SFR-DC 77 Y - safety-significant 78 Primary coolant system interfaces.
SFR-DC 78 Y - safety-significant, SARRDLs 79 Cover gas inventory maintenance.
SFR-DC 79 N
80 Reactor vessel and reactor system structural design basis.
MHTGR-DC 70 N
81 Reactor building design basis.
MHTGR-DC 71 Y - MHTGR-specific language 82 Provisions for periodic reactor building inspection.
MHTGR-DC 72 Y - MHTGR-specific language
A.Use of the term safety-significant
- Change: Replace important to safety from RG 1.232 DC with safety-significant to align with language from NEI 18-04
- DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor ApplicationsRoadmap (ML23277A139) identified that there may be some SSCs that may be important to safety but not safety-significant per NEI 18-04 process
- No gap because of use of RG 1.232 DC (e.g., those related to managing and monitoring effluents resulting from normal operations) 28
B. Use of graded approach to coolant boundary quality
- Change: Modified to indicate safety-significant elements of the primary or intermediate coolant boundary
- Consistent with NEI 18-04 approach, not all elements of primary coolant boundary are considered safety-related (SR) a priori
- Proper application of NEI 18-04 would appropriately classify structures, systems, and components (SSCs), resulting in quality, design, and performance requirements commensurate with safety significance
- SE notes that if primary coolant boundary components are not SR, an exemption may be needed from regulations 29
C. Use of SARRDLs
- Change: SARRDLs used instead of specified acceptable fuel design limits (SAFDLs)
- SARRDLs are compatible with Natrium design/licensing approach
- High-reliability metallic fuel chemically compatible with coolant
- Can establish fuel design limits as surrogates for SARRDLs
- SARRDLs are consistent with NEI 18-04 process that requires mechanistic source term evaluations
- SARRDLs provide appropriate performance-based approach to determining functional containment performance criteria
- Same basis as SARRDLs in RG 1.232
- No staff determination on specific SARRDLs 30
D. Use of functional containment concept
- Per previous discussion on SARRDLs, functional containment is also compatible with Natrium design and NEI 18-04 process
- Low-pressure operation
- Margin to coolant boiling
- Chemical compatibility between fuel and coolant
- Lack of sodium-water interaction
- No staff determination on specific functional containment barriers or performance 31
E. Other generic changes
- Change: Adoption of MHTGR-DC without MHTGR-specific language
- Change: Use of the term safe shutdown
- Sensitivity to cold shutdown for SFRs, coolant freezes at ambient temp
- Change is consistent with RG 1.232, SECY-94-084, and NEI 18-04/RG 1.233
- Change: Leak-tightness of cooling systems
- Anticipated in RG 1.232
- Natural draft air circulation system used for emergency core cooling
- Some amount of leakage not anticipated to impact ability of system to perform safety function 32