ML23031A060
| ML23031A060 | |
| Person / Time | |
|---|---|
| Issue date: | 01/12/2023 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-2228 | |
| Download: ML23031A060 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Kairos Power Licensing Subcommittee Open Session Docket Number:
(n/a)
Location:
teleconference Date:
Thursday, January 12, 2023 Work Order No.:
NRC-2228 Pages 1-112 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.
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1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
(ACRS) 5
+ + + + +
6 KAIROS POWER LICENSING SUBCOMMITTEE 7
+ + + + +
8 OPEN SESSION 9
+ + + + +
10 THURSDAY 11 JANUARY 12, 2023 12
+ + + + +
13 The Subcommittee met, via Teleconference, 14 at 9:30 a.m. EST, David A. Petti and Ronald G.
15 Ballinger, Chairs, presiding.
16 17 COMMITTEE MEMBERS:
18 DAVID A. PETTI, Chair 19 RONALD G. BALLINGER, Chair 20 VICKI M. BIER, Member 21 CHARLES H. BROWN, JR., Member 22 VESNA B. DIMITRIJEVIC, Member 23 WALTER L. KIRCHNER, Member 24 GREGORY H. HALNON, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 JOSE MARCH-LEUBA, Member 1
JOY L. REMPE, Member 2
MATTHEW W. SUNSERI, Member 3
4 ACRS CONSULTANTS:
5 DENNIS BLEY 6
8 DESIGNATED FEDERAL OFFICIALS:
9 WEIDONG WANG 10 CHRISTOPHER BROWN 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 CONTENTS (CONTINUED) 1 2
Kairos Power Topical Report (KP-TR-013), "Metallic 3
Materials Qualification for the Kairos Power Fluoride 4
Salt-Cooled High-Temperature Reactor" 5
6 Call to Order.................. 77 7
NRC Staff Introductory Remarks
......... 80 8
Richard Rivera, NRR............ 80 9
Metallic Materials Qualification Topical Report 10 (KP-TR-013)................... 81 11 John Price, Kairos Power
......... 81 12 George Young, Kairos Power
........ 85 13 Metallic Materials Qualification Topical Report 14 Safety Evaluation................ 95 15 Richard Rivera, NRR............ 95 16 John Honcharik, NRR............ 96 17 Public Comment (None).............
111 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
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3 4
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9 10 11 12 13 14 15 16 17 18 19 20 MEMBER BALLINGER: Okay. It's now a little after 2:00. The meeting will come to order.
This is a meeting of the Kairos Power Licensing Subcommittee of the Advisory Committee on Reactor Safeguards. I'm Ron Ballinger, techno-illiterate Chairman of today's Subcommittee meeting.
ACRS members in attendance are Charles Brown, Joy Rempe, Matt Sunseri -- keep going down the line here -- Vicki Bier, Dave Petti, Greg Halnon, Jose March-Leuba, and our Consultants Dennis Bley and Steve Schultz are also participating. And if I missed somebody, I apologize.
21 MEMBER KIRCHNER: Ron, Walt, I'm here.
22 MEMBER BALLINGER: Ah, okay, Walt.
23 MEMBER DIMITRIJEVIC: I am here, too.
24 MEMBER BALLINGER: And Vesna. Boy, I've 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 got two strikes on me already.
1 Okay. Chris Brown of the ACRS staff is 2
the Designated Federal Official for this meeting.
3 During today's meeting the Subcommittee 4
will review staff's Safety Evaluation on Topical 5
Report "Metallic Materials Qualification for the 6
Kairos Power Fluoride Salt-Cooled High-Temperature 7
Reactor," Revision 4.
8 The Subcommittee will hear presentations 9
by and hold discussions with the NRC staff, Kairos 10 Power representatives, and other interested persons 11 regarding this matter.
12 Part of the presentations by the Applicant 13 and the NRC staff may be closed in order to discuss 14 information that is proprietary to the licensee and 15 its contractors, pursuant to 5 USC 552b(c)(4).
16 Attendance at the meeting that deals with such 17 information will be limited to NRC and its 18 consultants, Kairos Power, and those individuals and 19 organizations who have entered into an appropriate 20 confidentiality agreement with them. Consequently, we 21 need to confirm that we have -- at that time, if we 22 need a closed session, we'll need to confirm that we 23 have only authorized people available, and we'll have 24 a different invitation for this.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 The rules for participation in all ACRS 1
meetings, including today's, were announced in The 2
Federal Register on June the 13th, 2019.
3 The ACRS section of the U.S. NRC public 4
website provides our Charter, Bylaws, agendas, Letter 5
Reports, and full transcripts of all full and 6
subcommittee meetings, including slides presented 7
here. The meeting notice and agenda for this meeting 8
were posted there.
9 We have received no written statements or 10 requests to make an oral statement from the public.
11 The Subcommittee will gather information, 12 analyze relevant issues and facts, and formulate 13 proposed positions and actions, as appropriate, for 14 deliberation by the full Committee.
15 The rules for participation in today's 16 meeting have been announced as part of the notice of 17 this meeting previously published in The Federal 18 Register.
19 A transcript of the meeting is being kept 20 and will be available, as stated in The Federal 21 Register notice.
22 Due to the COVID pandemic, today's meeting 23 is being held over Microsoft Teams for the ACRS, NRC 24 staff, and the licensee's attendees. There is also 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
80 telephone bridge line allowing participation of the 1
public over the phone.
2 When addressing the Subcommittee, the 3
participants should, first, identify themselves and 4
speak with sufficient clarity and volume, so that they 5
may be readily heard. When not speaking, we request 6
that participants should mute your computer 7
microphone, or phone by pressing *6, if you've had it 8
on.
9 I'll now proceed with the meeting, and I'd 10 like to call on the NRC staff. I'm not sure who is 11 going to make a comment for the NRC. If there's a 12 member that wants to make a comment, please do so.
13 MR. RIVERA: Good afternoon. This is 14 Richard Rivera, Quality Manager for the review of this 15 Topical Report.
16 The remarks that were covered by my Branch 17 Chief, William Jessup, that were made this morning, 18 also cover the discussion of this Topical Report. So, 19 we have no additional remarks.
20 MEMBER BALLINGER: Thank you.
21 MR. RIVERA: Thank you.
22 Okay. So, first up -- did I hear another 23 person that might want to make a comment?
24 (No response.)
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 Okay. Thank you.
1 Okay. So, Kairos, I guess it's George 2
Young. Are you up?
3 MR. PRICE: Hello. This is John Price.
4 Can you hear me okay?
5 MEMBER BALLINGER: Ah, sorry about that.
6 Yes, I can.
7 MR. PRICE: Okay. Okay. We'll go ahead 8
and get started.
9 Hello. My name is John Price. I'm a 10 Senior Licensing Engineer for Kairos Power, and I'll 11 be presenting with Dr. George Young, on the Metallic 12 Materials Qualification Licensing Topical Report, 13 Revision 4.
14 I have a bachelor's degree in nuclear 15 engineering from Texas A&M University, and I'm proud 16 to say that I've been in the regulatory part of this 17 industry for over 40 years, in advanced reactor and 18 SMR design for the past 10 years, and with Kairos 19 Power for the past four years.
20 As we always start off, our company's 21 vision is to enable the world's transition to clean 22 energy, with the ultimate goal of dramatically 23 improving people's qualify of life while protecting 24 the environment. What this means as we go through 25 NEAL 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 this meeting is that, as the Licensing Topical Report 1
gets approval from the ACRS, we are one step closer to 2
transitioning to cleaner energy and improving people's 3
qualify of life while protecting the environment.
4 Today, Dr. Young and I would like to 5
present the Metallic Materials Qualification Topical 6
Report and our methods used to qualify these materials 7
for use in the KP-FHR, specifically addressing the 8
environmental effects on materials.
9 The testing plan is for metallic materials 10 used in Flibe-wetted areas for safety-related, high-11 temperature components in the non-power test reactor 12 which we call Hermes, and for the commercial power 13 reactor which we'll designate as the KPX reactor.
14 The Topical Report uses a phenomena 15 identification and ranking table, the PIRT test plan, 16 as a baseline for the KP-FHR -- this is from Reg Guide 17 1.203 -- used to identify significant degradation 18 phenomena and to develop the testing and modeling 19 qualification presented in the Topical Report. As 20 this is a methodology document, the demonstration and 21 qualification will be documented in the Safety 22 Analysis Reports as part of our future licensing 23 action.
24 Any questions on this?
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83 MEMBER KIRCHNER: Yes. Hi. This is Walt 1
Kirchner.
2 Just a general question. You talked about 3
the Flibe-wetted areas of the primary system. Are you 4
considering other materials for internals, and you 5
would address those in a separate TR? Or is your 6
intent, by and large, to use the two materials 7
identified here as internals for the reactor vessel 8
and primary coolant system?
9 MR. PRICE: Yes. So, the 316H materials 10 will be used for the safety-related areas for the 11 Hermes and the KPX. We may get into using other 12 materials for non-safety-related areas. But, for the 13 safety-related areas, these are the materials that 14 we'd be using there, covered by this report.
15 Does that answer your question?
16 MEMBER KIRCHNER: Yes, it does. Thank you 17 very much. I was just curious whether you were 18 considering Inconels or other materials for things 19 like control rods and such. Thank you.
20 MR. PRICE: So, this slide right here is 21 a depiction of table 1 from the Metallic Materials 22 Topical Reports and provides a summary of key 23 parameters for the commercial power reactor, as well 24 as the non-power test reactor.
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84 The main thing to note on here is the 1
lifetime of the non-power test reactor will only be 2
one year commissioning and four years of operation.
3 For this reason, as we'll discuss later on, there will 4
be a reduced test duration for the shorter-lived non-5 power test reactor.
6 Slide 5 represents the organization of the 7
Licensing Topical Report. From this, I'd like to say 8
that this test plan and analysis ensures extremely low 9
probability of abnormal leakage, rapidly propagating 10 failure, or gross rupture, which partly satisfies the 11 principal design criteria of PDC 14, the Reactor 12 Coolant Boundary, and PDC 31, which is the Fracture 13 Prevention of the Reactor Coolant Boundary.
14 At this time, I'm going to turn it over to 15 Dr. George Young to complete the open session 16 presentation. Dr. Young is a fellow scientist at 17 Kairos Power and has a BS in materials engineering 18 from Renssalaer Polytechnic Institute and an MS and 19 PhD degrees in material science from the University of 20 Virginia. He has over 30 years of experience in the 21 nuclear power industry and is expert with material 22 selection and performance for both conventional and 23 advanced nuclear power systems.
24 At Kairos Power, Dr. Young leads the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 structural material qualification effort and 1
environmental degradation testing. Dr. Young has 2
authored over 50 peer-reviewed articles and 3
books/chapters in the research area of 4
environmentally-assisted cracking, weld metallurgy, 5
and physical metallurgy.
6 So, Dr. Young, take it away.
7 DR. YOUNG: Thanks, John.
8 So, in Chapter 2 of the Topical Report, we 9
talk about why we chose 316H and the weld filler metal 10 16-8-2. These materials are currently qualified in 11 the ASME Code for high-temperature applications, and 12 as material is qualified, then components can be 13 designed, as John mentioned, for extremely low 14 probability of abnormal leakage and resistance to 15 rapidly propagating failure, as well as resistance to 16 gross rupture.
17 We know from its tremendous industrial 18 use, including use in nuclear power systems, 316H and 19 its weld metals exhibit desirable mechanical 20 properties, and they have some -- and growing, I'll 21 say -- demonstrated capability and compatibility with 22 Flibe salt and an extensive experience base.
23 Also, I'll just note that 316H and its 24 weld materials are used in other critical industries, 25 NEAL 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 like oil and gas and the chemical process industry, in 1
critical applications at the times and temperatures 2
relevant that we're designing components with for the 3
KP-FHR.
4 CHAIR PETTI: George?
5 DR. YOUNG: Yes?
6 CHAIR PETTI: This is Dave Petti. I just 7
have a question. I'm trying to understand.
8 I've sat through more presentations than 9
I can remember where experts from Oak Ridge said that 10 316 corroded too much in salt, and that was why Oak 11 Ridge developed Astralloy X, you know, after the MSRE 12 experience.
13 Now, I understand that Astralloy X isn't 14 Code-qualified, but where is the disconnect? You 15 guys, obviously, think 316H is acceptable. Is it an 16 issue of just lifetime, how to justify the two views?
17 DR. YOUNG: You know, I've thought about 18 this, also, Dave. And Oak Ridge published some 19 corrosion data for 316, and it wasn't too bad. And I 20 always wondered if maybe just they had the resources 21 to do long-term alloy development and wanted to do 22 that. But we see no limitation corrosion, or 23 otherwise, with using 316H. And we'll talk about some 24 of those data.
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87 CHAIR PETTI: Yes. No, it's a lot cheaper 1
than Astralloy.
2 DR. YOUNG: Yes.
3 CHAIR PETTI: But, yes, okay. I mean, I'm 4
glad you came to the same conclusion because that 5
disconnect has always been in the back of my mind.
6 So, thanks.
7 DR. YOUNG: Yes. I mean, I don't know why 8
Oak Ridge did it. If I could do alloy development, 9
that's fun work, but it's not really needed.
10 CHAIR PETTI: Yes, yes. Okay. Thanks.
11 DR. YOUNG: And then, in Chapter 3.1, we 12 talk about testing required for the ASME Code 13 extension, and we note that the base metal 316H is 14 qualified up to 816 C, but the weld filler metal that 15 we've chosen to match the properties there, 16A2, 16 that's currently only qualified to 650 C. So, Kairos 17 Power has taken on this Code extension, and we will 18 fully test that -- tensile test, creep fatigue, and 19 creep rupture testing -- to get that to match the base 20 metal up to 816 C.
21 3.2 talks about testing now to facilitate 22 both the Hermes and the commercial power reactor 23 designs. These are tests being conducted for the 24 stainless steels, for the ASME high-temperature 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
88 code model calibration and validation. And these are 1
all the tests we're conducting in high-temperature air 2
to do that. And those include tensile testing, stress 3
relaxation testing, exchange rate change or sometimes 4
called stress dip tests, uniaxial creep testing, as 5
well as creep fatigue testing.
6 And then, 3.3, some more high-temperature 7
air testing. And just a couple of notes. Based on 8
good feedback from the NRC, we've added a testing 9
program to look at stress through oxidation cracking 10 of 316H heat-affected zones relative to an alloy 11 that's known to be susceptible, specifically, alloy 12 347.
13 And then, as we'll talk about in a little 14 bit more detail, things like thermal stresses and 15 thermal striping were brought up in the metallics PIRT 16 as a potential concern. Those are currently managed 17 by design methods, and we feel that there's no further 18 testing needed to take those potential issues off the 19 plate.
20 Next slide.
21 And then, maybe really the meat of the 22 report is on environmental compatibility, and we talk 23 about the use of the PIRT data and the PIRT process.
24 I think we had a fantastic PIRT and it took a somewhat 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
1 2
3 4
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9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 89 conservative view toward what we would assess. And we'll talk about that.
The PIRT phenomena are grouped into four major categories: corrosion, environmentally-assisted cracking, kind of metallurgical effects or "other" section, and finally, irradiation effects. And I'll talk about this in a little bit more detail as we go to the tables in a couple of slides.
Just to highlight, we've had a lot of very good, very technical exchange with the NRC staff.
Some of the major comments that have been addressed are as follows:
We've added a lot of heat-affected zone samples throughout the EAC testing and corrosion testing. And these samples are kind of prioritized for the non-power Hermes reactor.
As I
mentioned, we added a
pretty significant direct testing program for the concern over stress relaxation cracking, and we're going to do a comparison study between 347 heat-affected zones, where it's of known concern, and a 316H sample set, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
90 where it's a potential concern.
1 And then, we've also added a lot of detail 2
to the testing plans -- numbers of samples, 3
replicates, how we treat the data, exposure times, et 4
cetera.
5 CHAIR PETTI: George?
6 DR. YOUNG: Yes?
7 CHAIR PETTI: This is Dave again.
8 Just another question on the scope. I 9
understand this is good for the vessel and the piping.
10 But, like the control logs, do they have some sort of 11 metallic enclosure that they sit in? Is that sort of 12 outside the scope of this, or how does that --
13 DR. YOUNG: Yes, so we focused on the 14 vessel as the main safety-related component to contain 15 fission products.
16 And, John, do you want to address our 17 thoughts on how we're going to address concerns with 18 the control rods? We think we're going to treat that 19 separately.
20 MR. PRICE: Yes, it will be treated 21 separately.
22 CHAIR PETTI: Okay. No, that's good.
23 Just so I know where the scrim line is.
24 MEMBER KIRCHNER: Dave, that was my 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 earlier question.
1 CHAIR PETTI: That's what I thought.
2 MEMBER KIRCHNER: Yes.
3 CHAIR PETTI: I just was going to follow 4
up just to make it clear at least in my mind.
5 MEMBER KIRCHNER: Yes.
6 CHAIR PETTI: Because there you could get 7
more than.1 BPA on the materials and it becomes more 8
complex then.
9 MEMBER KIRCHNER: Right. Yes.
10 CHAIR PETTI: Okay. Thanks.
11 MEMBER BALLINGER: This is Ron Ballinger.
12 Maybe this is for the closed session, but 13 I understand, and I see data on 321 and 347 on 14 cracking, but I haven't seen any data on 316H on that 15 type of cracking. Have you seen it?
16 DR. YOUNG: The AGR, the gas reactor 17 experience, there is some stress relaxation cracking 18 in 316H heat-affected zones. As best as I can tell, 19 it's in fairly atypical heats, high-sulfur heats, 20 somewhat abused material where you get significant 21 strain hardening from
- welding, probably some 22 segregation. And then, all those factors stack up.
23 And the British have seen stress relaxation cracking 24 in some 316H.
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92 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 MEMBER BALLINGER: Okay. Thanks.
DR. YOUNG: All right. So, just to summarize, we've selected 316H base metal and ER16-8-2 weld filler metal as the metallic structural alloys for use in our safety-significant, high-temperature, Flibe-wetted components.
We're doing testing to support licensing of both our non-power test reactor Hermes -- that's got a four-year, full-power life, so kind of four-plus-one hot commissioning -- and the commercial power reactor with a xxxxxxxxxxxxxx KPX.
We're focusing the testing on these two alloys for the reactor vessel. And again, that's our primary safety-related component.
As you've noted from some of these schematics, most all the penetrations are up in the head. We don't have any low penetrations in the vessel.
We're going to both extend the Code qualifications for weld metal 16-8-2 to match the base metal, and then, the materials testing effort to look at potential environmental effects involves high-temperature air testing for ASME design, and then, testing in Flibe salt to account for any corrosion, stress corrosion, et cetera.
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93 Any further questions on that?
1 CHAIR PETTI: Yes.
2 DR. YOUNG: Yes?
3 CHAIR PETTI: Just one more question. The 4
Topical talks about that you may clad certain areas 5
for enhanced corrosion resistance. Is that just sort 6
of risk-reduction activity that Kairos is pursuing, 7
you know, in the event that the testing doesn't go as 8
you anticipate, based on all the data that you've got 9
in the Topical thus far?
10 DR. YOUNG: Exactly. We thought for years 11 ago that we should leave open that possibility. We 12 had a little GAIN project that looked at cladding.
13 But, as we'll touch on in the closed session, we don't 14 have any plans or any need to --
15 CHAIR PETTI: Okay. I mean, I have sort 16 of a Code question. Were you to clad stuff, does that 17 remove the need for environmental testing because your 18 structure doesn't see the coolant?
19 DR. YOUNG: I don't think so. I think 20 those clad rules are drafted and being developed. And 21 so, there's a compliant clad and hard clad, and 22 certainly some benefit you can take there. I don't 23 think you could get away with no --
24 CHAIR PETTI: No testing? Yes.
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94 MEMBER BALLINGER: But I can answer that.
1 No, you can't.
2 MEMBER KIRCHNER: You can, yes.
3 MEMBER BALLINGER: No.
4 MEMBER KIRCHNER: The BWRs --
5 MEMBER BALLINGER: You have to assume that 6
the cladding has a breach.
7 CHAIR PETTI: Ah, okay.
8 MEMBER BALLINGER: So, you can't get away 9
with that.
10 CHAIR PETTI: Yes. Okay. Thanks.
11 MEMBER BALLINGER: And, you know, from my 12 perspective, the fact that you're going to qualify 13 these materials in all respects to Division 5, to 14 Section III, Division 5, and everything proceeds from 15 that, right?
16 DR. YOUNG: Right.
17 Any other questions or thoughts?
18 MEMBER KIRCHNER: George?
19 DR. YOUNG: Yes.
20 MEMBER KIRCHNER: This is Walt Kirchner 21 again.
22 Just could you address in the closed 23 session fasteners and other internals?
24 DR. YOUNG: Sure.
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95 MEMBER BALLINGER: Okay. Is that the 1
extent of the Kairos presentation?
2 DR. YOUNG: Yes, it is.
3 MEMBER BALLINGER: In the non-closed 4
session? Excuse me.
5 DR. YOUNG: Yes.
6 MR. PRICE: Yes.
7 MEMBER BALLINGER: Okay. Are there any 8
questions for Kairos from the members or consultants 9
before we switch over to the staff?
10 (No response.)
11 Hearing none, okay. We are proceeding at 12 a rapid rate. Can we get the staff presentation up?
13 MR. RIVERA: Yes. This is Richard Rivera, 14 Project Manager for this Topical Report. Give me one 15 moment to switch to the presentation.
16 So, can someone confirm if you can see the 17 slides?
18 MEMBER BALLINGER: Very well. Thank you.
19 MR. RIVERA: Excellent.
20 So, I will pass on the mic to John 21 Honcharik, who is the lead reviewer for this Topical 22 Report.
23 MR.
HONCHARIK:
All right.
Good 24 afternoon.
25 NEAL 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 I'm John Honcharik, Senior Materials 1
Engineer in the Division of New and Renewed Licensing.
2 Alex Chereskin is in the Division of Advanced New 3
Reactors. And both of us reviewed the Topical Report 4
for Metallics Qualification and we'll present to you 5
our evaluation and conclusion of that Topical Report.
6 Next slide.
7 So, the NRC staff has reviewed the Topical 8
Report which provides the qualification plan for 9
metallic structural materials used in Flibe-wetted 10 areas for the safety-related, high-temperature 11 components of the KP-FHR power and test reactors.
12 The planned material testing, including 13 analysis and monitoring programs, will be used to 14 address the material reliability and compatibility of 15 the metallic materials in an environment of these 16 designs.
17 The results of these planned testing and 18 analysis will be used in future licensing applications 19 and provide a detailed description of the design, 20 inspection, and surveillance programs for the KP-FHR 21 design in order to demonstrate the materials' 22 reliability.
23 Next slide.
24 So, these are the regulations, and they 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 describe what information would be in an application, 1
such as a construction permit, which would include the 2
materials and its compatibility to the design 3
environment, as described in this Topical Report.
4
- Also, the Topical Report KP-TR-003 5
provides the NRC-approved principal design criteria 6
for the KP-FHR design. The PDCs below are applicable 7
to the qualification of the metallic components for 8
the KP-FHR design which the staff based this review 9
on. The two are PDC 14 and 31, which, basically, 10 relate to ensure that the reactor coolant pressure 11 boundary have an extremely low probability of abnormal 12 leakage, rapidly propagating failure, and to fail in 13 a non-brittle manner, and to minimize the probability 14 of rapidly propagating failure.
15 Next slide.
16 The staff's review focused on the overall 17 testing framework, to include if there is reasonable 18 assurance of the testing for environmental effects of 19 Flibe on the metallic structural materials provided in 20 Section 4 of this Topical Report, and that will also 21 be discussed today, which are: the material, which 22 are 316H and its fill metal ER-16-8-2; the test 23 environment as it relates to these designs, and the 24 four degradation mechanism categories of corrosion, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
98 environmentally-assisted
- cracking, effects of 1
metallurgical properties, and irradiation.
2 Next slide.
3 Within the four degradation categories 4
noted previously, 14 types of degradation mechanisms 5
were proposed to be evaluated by testing, analysis, or 6
surveillance. How these degradation mechanisms were 7
addressed in the qualification program will be 8
discussed later in the presentation.
9 Next slide.
10 First, we'll discuss the material to be 11 used. The metallic structural materials proposed for 12 the KP-FHR designs are 316H, also known as stainless 13 steel, and the associated stainless steel weld fill 14 metal ER-16-8-2. These materials are qualified for 15 use in ASME Code,Section III, Division 5, for high-16 temperature reactors with respect to their mechanical 17 properties. Division 5 of Section III will be 18 endorsed in Reg Guide 1.87, Revision 2, and will be 19 published soon.
20 There is Condition 4 related to the weld 21 material mechanical properties, as George alluded to 22 earlier, since the fill metal is not currently 23 qualified to the higher temperatures necessary to 24 support the accident scenarios of these designs. So, 25 NEAL 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 therefore, the NRC imposed the condition that the fill 1
metal must be qualified to that temperature, in 2
accordance with the requirements of Division 5 that 3
balance the postulated accident conditions, and 4
approved by the NRC.
5 The staff's review of the Topical Report, 6
as presented
- today, will only address the 7
environmental effects of Flibe on the metallic 8
material, since the mechanical properties are already 9
covered in Division 5 for the ASME Code.
10 Next slide.
11 Next, we will discuss the test environment 12 used for the proposed material testing. The staff 13 found that the material testing environment duplicates 14 the operating environment of the KP-FHR designs listed 15 above. Higher test temperatures are used in some 16 tests in order to develop environmental degradation 17 rates during postulated accident scenarios.
18 Testing conducted in nominal Flibe which 19 is applied with minimal water and contaminants, the 20 testing in this nominal Flibe will provide more 21 conservative results since the nominal Flibe is more 22 oxidizing and will produce greater corrosion rates.
23 Additional testing will be performed in redox 24 controlled Flibe to see if these corrosion rates 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
100 be minimized.
1 There is Condition and Limitation No. 3 2
related to the environment to be tested. The staff 3
imposed a condition that, if the time and temperature 4
for both normal operation and postulated accident 5
conditions change for these designs, they must still 6
be bounded by the NRC staff-endorsed ranges found in 7
table 2 of Reg Guide 1.87.
8 Next slide.
9 Concerning the postulated accident 10 scenarios, there are two potential accident scenarios 11 for the commercial power reactor. Those are 12 intermediate salt ingress and, also, air ingress into 13 the Flibe. For the non-power test reactor, there is 14 one potential accident scenario, which is an air 15 ingress into the Flibe.
16 These impurity ingresses would produce a 17 specific concentration of these impurities that could 18 affect a safety-related component. So, the NRC finds 19 this approach of testing with these impurities that 20 simulate these accident conditions acceptable.
21 For the power reactor, testing with both 22 air and intermediate salt as impurities is reasonable 23 to develop the effects of impurities on the corrosion 24 rate on 316H and its fill metal because it will bound 25 NEAL 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 the accident conditions for the power reactor.
1 Likewise, for the test reactor, it will be done with 2
air as an impurity, which we found acceptable also.
3 As part of this testing, Condition 3, which I stated 4
previously in a previous slide, would be applicable.
5 In addition, Condition 8 would apply, 6
since the details of this impurity testing, that is, 7
the concentration of each of the contaminants, have 8
not been determined. Therefore, the specific 9
concentration of each contaminant used in impurity 10 effects testing on 316H and its fill metal shall bound 11 the accident scenarios postulated in the transient 12 analysis for these KP-FHR designs.
13 The next topic reviewed for the Topical 14 Report will be the degradation mechanisms. The 15 following slides will discuss each degradation 16 mechanism accounted for in the qualification program.
17 The first is corrosion. This includes the 18 various types of corrosion, such as general corrosion, 19 crevice corrosion, thermal aging, and erosion and 20 wear. The Topical Report proposes that the corrosion 21 testing will be used to develop quantitative corrosion 22 models for 316H and its fill metal in the Flibe 23 environment.
24 The proposed testing will use 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
102 representative coupons of these materials and 1
conditions described in tables 12 and 13 of the 2
Topical Report, and will be performed under different 3
conditions to assess the impact of these specific 4
corrosion degradation mechanisms.
5 Next slide.
6 So, the NRC found the proposed testing 7
acceptable to determine the impact Flibe has on the 8
corrosion rates of the material, based on the 9
following: the impact of temperature; microstructure; 10 the salt composition, which would include the nominal 11 redox and impurity ingresses; the geometry; 12 erosion/corrosion, thermal aging, and presence of 13 graphite and redox control on the corrosion rate of 14 316H and its fill metal, which can be determined from 15 these proposed tests.
16 For example, the NRC finds the proposed 17 tests acceptable for simulating the occlusal or 18 crevice geometry effects in order to determine whether 19 or not crevice corrosion is a concern for these 20 materials in Flibe.
21 Tests for the effect of temperature on the 22 corrosion rate would also be acceptable because the 23 corrosion is evaluated over the range of temperatures, 24 consistent with both operating temperatures 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
103 postulated bounding accident conditions.
1 Next slide.
2 Continued acceptability of the impacts of 3
corrosion. Tests are consistent with the expected 4
corrosion mechanisms for 316H in a molten salt 5
environment. In addition, some tests will be 6
conducted with foam Flibe, which is necessary in order 7
to simulate the flowing salt in the reactor.
8 Graphite particles will be used in tests 9
with flow in order to determine the effects of these 10 particles on rates of erosion/corrosion.
11 In addition, the effects caused by 12 microstructure changes due to weld, heat-affected 13 zones, thermal aging, and cold working on the 14 corrosion rates of the material will also be tested 15 and evaluated.
16 Next slide.
17 The next degradation category is for 18 environmentally-assisted cracking. This includes 19 stress corrosion cracking, corrosion fatigue, and 20 environmental creep.
21 For stress corrosion cracking and 22 corrosion fatigue, the NRC finds the proposed test 23 plan provides reasonable assurance in determining the 24 crack growth rates for these two mechanisms 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
104 environment of the KP-FHR designs, because in situ 1
mechanical testing will be performed using slow strain 2
rate testing for corrosion fatigue and stress 3
corrosion cracking in both normal and accident 4
conditions. The SSRT testing will be conducted in 5
nominal Flibe and redox controlled Flibe to assess if 6
316H and its fill metal are susceptible to stress 7
corrosion cracking. The samples to be tested are 8
representative of the material and weldments to be 9
used in the KP-FHR designs, and therefore, are 10 acceptable. The testing methodology will also be 11 conducted in accordance with well-established ASTM 12 specifications such as G129.
13 Next.
14 For environmental creep degradation, the 15 NRC found the proposed testing plan acceptable because 16 pretesting in both nominal Flibe and in air using 17 welded base metal samples, to include both base metal, 18 weld metal, and heat-affected zones, ensures each 19 material's state is tested in the operating 20 environment.
21 Creep tests will be performed in both air 22 and in nominal Flibe. Tests in nominal Flibe will 23 determine if the Flibe contributes any additional 24 degradation beyond those determined from the creep 25 NEAL 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 tests performed in air. Additional testing will be 1
required to quantify any increase in degradation 2
contributing to the Flibe if the testing determines 3
Flibe has this additional effect.
4 Next slide.
5 The next degradation category is for 6
metallurgical effects, as designated in the Topical 7
Report. This includes the degradation mechanism of 8
stress relaxation
- cracking, phase formation 9
embrittlement, and thermal cycling.
10 For stress relaxation cracking, the NRC 11 found the proposed testing plan acceptable. Both the 12 current data and experience shows that only the heat-13 affected zone of the 316H base metal with high 14 triaxial stresses are susceptible. Testing in air is 15 acceptable since the triaxial stresses are the major 16 contributors to the stress relaxation cracking and not 17 the environment.
18 Comparing the susceptibility of 316H to 19 that of 347, which is known to be susceptible, would 20 allow a determination of the bounding triaxial 21 stresses.
22 Test results will be used, then, to 23 conduct further analysis and design requirements of 24 the KP-FHR design, such as weld designs and specific 25 NEAL 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 weld processes and parameters, in order to reduce 1
these triaxial stresses, and thereby, minimizing 2
stress relaxation cracking.
3 Next slide.
4 For phase formation embrittlement, the NRC 5
found the proposed testing plan acceptable because 6
phase formation embrittlement may occur when 316H and 7
its fill metal picks up an element during its exposure 8
to Flibe and forms a deleterious second phase.
9 Testing will determine whether this degradation 10 mechanism occurs for the KP-FHR design environment.
11 As part of this testing, Condition 11 is 12 applicable, which states that, if intermetallic 13 formation occurs, an applicant will need to perform 14 testing to quantify the effects on the mechanical 15 properties of 316H and its fill metal.
16 For thermal cycling --
17 MEMBER KIRCHNER: Richard, this is Walt.
18 Just historically, what elements would 19 contribute to that last phase embrittlement 20 phenomenon? When you say, "elements," are talking 21 trace fission products, erosion or corrosion products 22 from the metals?
23 MR. HONCHARIK: Yes, this is John 24 Honcharik.
25 NEAL 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 I think right now, for the open session, 1
I'm not sure I could say any more. We'd probably have 2
to answer that in the closed session. So, I can make 3
a note of that.
4 MEMBER KIRCHNER: That's fine. Thank you.
5 MR. HONCHARIK: Yes.
6 Okay. So for thermal cycling, the NRC 7
staff found the proposed testing plan acceptable 8
because large thermal transients could lead to high 9
stresses, resulting in thermal fatigue degradation of 10 316H and its fill metal. These thermal stresses will 11 be addressed by conducting analysis in order to refine 12 the design and operation of the KP-FHR designs to 13 mitigate these large thermal gradients.
14 As part of this analysis, Condition and 15 Limitation 12 apply and the applicant will assess the 16 thermal cycling and striping in future licensing 17 submittals by minimizing the thermal gradients via 18 appropriate design and operating conditions.
19 Next slide.
20 The next degradation category is for 21 irradiation-induced embrittlement. This includes 22 irradiation-induced embrittlement, irradiation-23 affected corrosion, and irradiation-assisted stress 24 corrosion cracking.
25 NEAL 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 For irradiation-induced embrittlement, the 1
NRC staff found the proposed testing plan acceptable 2
because the data shows that irradiation-induced 3
embrittlement at low strain rates can affect the 4
material properties, such as tensile strength, (audio 5
interference), and creep (audio interference) due to 6
generation healing.
7 Degradation factors will be developed 8
based on these existing data. However, irradiation 9
tests will be conducted on 316H and its fill metal, 10 including the heat-affected zone of the 316H, in order 11 to quantify any design margins at the irradiation 12 levels posed for both the test reactor and the 13 commercial reactor. The details of that testing will 14 be provided in a future application.
15 As part of this testing, Limitation and 16 Condition 13 would apply and that the test environment 17 shall bound the KP-FHR designs, including expected 18 irradiation damage and healing content.
19 For irradiation-affected corrosion, the 20 NRC staff found the proposed testing plan acceptable 21 because existing data shows that irradiation may 22 increase general corrosion rates. However, the 23 materials surveillance program for the non-power test 24 reactor and commercial power reactor will be used 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
109 monitor irradiation-affected corrosion. In addition, 1
an inspection and monitoring program will assess the 2
wall thickness of the reactor vessel and will be 3
implemented. So, therefore, existing data will be 4
used in order to develop the degradation factors and 5
be monitored by the surveillance program.
6 And as part of that surveillance program, 7
Condition 14 would apply, and that this program, 8
including the inspection and monitoring program, must 9
be implemented for the test reactor and commercial 10 power reactor. You will use the KP-FHR designs to 11 assess and monitor both irradiation-affected corrosion 12 rates and, also, for stress corrosion cracking, which 13 we'll talk about in the next slide.
14 So, for irradiation-assisted stress 15 corrosion cracking, NRC found the proposed testing 16 plan acceptable because irradiation-induced stress 17 corrosion cracking is not expected at the low 18 irradiation level. The stress corrosion cracking test 19 program specified in Section 4.24 will determine if 20 stress corrosion cracking is a credible degradation 21 mechanism for the environment in these designs.
22 Therefore, the material surveillance 23 program for both the non-power test reactor and the 24 commercial power reactor will be used in order 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
110 monitor this degradation mechanism. And the condition 1
that I stated previously would also apply to this 2
surveillance program.
3 So, in conclusion, the NRC staff finds 4
that there's reasonable assurance that the material 5
testing plan, including the analysis, surveillance, 6
and monitoring, for both 316H and its associated fill 7
metal, with the limitations and conditions noted in 8
the Safety Evaluation, can be used to provide the 9
necessary information to address the materials 10 reliability and compatibility in the environment of 11 the KP-FHR designs.
12 And that's based on, in summary:
13 The testing duplicates the operating 14 environment for both normal and accident conditions.
15 The material test samples are 16 representative of actual weldments.
17 Analysis will be performed to mitigate 18 stress relaxation cracking and thermal cycling through 19 design and operations.
20 The material surveillance program will be 21 used to monitor for irradiation effects on corrosion 22 and stress corrosion cracking.
23 And the results of the planned tests and 24 analysis will be used for future licensing 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
111 applications to ensure a component's performance 1
safety function and that there would be an extremely 2
low probability of abnormal leakage propagating 3
failure or brittle failure, which would partially 4
PDCs 14 and 31.
5 And I have it open for questions.
6 MEMBER BALLINGER: Questions from the 7
members and/or consultant?
8 (No response.)
9 Hearing none, before we switch to the 10 closed session, we'll open the line and ask for public 11 comment.
12 If there are members of the public that 13 would wish to make a comment, please unmute yourself 14 or press *6, and make your comment.
15 (No response.)
16 Hearing none, thank you very much for your 17 presentation.
18 We've been going very fast. So, I would 19 suggest that we take a break until 3:00 p.m., and 20 then, come back on in the closed session. So, we'll 21 take what amounts to about a 10-minute break, and 22 then, we'll see you in the closed session at 3:00 p.m.
23 (Whereupon, the above-entitled matter went 24 off the record at 2:52 p.m.)
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
KP-NRC-2301-001 Presentation Materials for the January 12, 2023, ACRS Kairos Power Subcommittee Meeting (Non-Proprietary)
Copyright © 2022 Kairos Power LLC. All Rights Reserved.
No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
Metallic Materials Qualification Topical Report ACRS Kairos Power Subcommittee Meeting January 12, 2023 OPEN SESSION
Proprietary Information Withhold from Public Disclosure under 10 CFR 2.390(a)(4)
Confidential and Proprietary, Part 810 Export-Controlled Information No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC Kairos Powers mission is to enable the worlds transition to clean energy, with the ultimate goal of dramatically improving peoples quality of life while protecting the environment.
Copyright © 2022 Kairos Power LLC. All Rights Reserved.
No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
3 3
Introduction Identify Standard Requirements Evaluate Existing Data Identify Data Gaps Identify Methods to Close Gaps Conduct Testing Complete Design Demonstrate Qualification in a Safety Analysis Report
- This report presents the methods for qualifying metallic materials for use in KP-FHRs Qualification is subject to the conditions in topical report Uses KP-X PIRT test plan as a baseline but with reduced test durations for the shorter-lived Non-Power Test Reactor
- Demonstration of qualification will be documented in safety analysis report documents as part of licensing applications under 10 CFR Part 50 or Part 52
- This report is applicable to the Non-Power Test Reactor (Hermes) and to the Commercial Power (KPX) Reactor
Copyright © 2022 Kairos Power LLC. All Rights Reserved.
No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
4 Summary of Key Parameters in KP-FHR (Table 1)
Parameter Power Reactor Non-Power Test Reactor Reactor Description Low pressure, fluoride salt-cooled high temperature reactor (FHR)
Core Configuration Pebble bed core, graphite reflector, and enriched Flibe molten salt coolant Physical Dimensions Reactor Vessel is ~4 m diameter,
~6 m height Reactor Vessel is ~2.5 m diameter,
~4 m height Reactor Thermal Power 320 MWth 35 MWth Primary Heat Transport System Flibe Salt, 550°C-650C, ~0.2 MPa, ~0.11-0.15 m/s Intermediate Heat Transport System Intermediate Coolant, <0.2 MPa, 360°C-600°C None. Primary Heat Transport System rejects heat to the (air) heat rejection subsystem Power Conversion System 300°C-585°C, steam ~19 MPa None. Primary Heat Transport System rejects heat to the (air) heat rejection subsystem Material for Safety-Related Structures Alloy 316H and ER16-8-2 (ASME Section III, Division 5, approved)
Lifetime
(( ))
5 years (1 year commissioning + 4 years operation)
End of Life Irradiation of Reactor Vessel
<0.1 dpa
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5 Metallic Materials Topical Report Organization
- Introduction KP-FHR Technologies Regulatory Information
- Structural Alloys
Background
Structural Alloy Selection Industrial Experience
- Air Testing and Finite Element Analysis Testing Required for ASME Code Extension Testing to Facilitate Reactor Designs High Temperature Testing & Analysis to Support Potential Degradation
- Compatibility with Flibe and Irradiation
- Environmental Compatibility
- Conclusions and Limitations
- Appendix B: Inspection and Aging Management
- Appendix C: Details of the Corrosion Data Analysis
- Appendix D and E: Not Used
- Appendix F Certified Material Reports
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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
6 Chapter 2: Alloy 316H and Weld Filler Metal ER16-8-2
- Alloy 316H and Weld Filler Metal ER16-8-2 were chosen because of its qualification in high temperature applications. As the material is qualified, components can be designed for extremely low probability of abnormal leakage, resistance to rapidly propagating failure, and resistance to gross rupture.
- Alloy 316H and its weld metals exhibit desirable mechanical properties, have demonstrated compatibility with Flibe salt, and have an extensive experience base in nuclear applications.
- Alloy 316H and its weld metals are used in other industry applications near the time and temperature of the KP-FHR.
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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
7 Chapter 3.1: Testing Required for ASME Code Extension
- Alloy 316H is qualified for 816°C and Weld Filler Metal is currently qualified to 650°C
- The weld filler metal will be tested to extend the ASME qualification of ER16-8-2 to match the base metal.
Elevated Temperature Tensile Testing Creep-Fatigue Testing Creep-Rupture Testing
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8 Chapter 3.2: Testing to Facilitate Non-Power and Commercial Power Reactor Designs
- The types of tests being conducted for Alloy 316H stainless steel model calibration and validation of ASME design methodologies (all conducted in high temperature air)
Tensile Testing Stress Relaxation Testing Strain Rate Change (Stress Dip) Testing Uniaxial Creep Testing Creep-Fatigue Testing
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9 Chapter 3.3: High Temperature Testing & Analysis to Support Potential Degradation
- Stress Relaxation Cracking - testing program per discussion with NRC Staff
- Thermal Stresses & Thermal Striping - managed via design
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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
10 Chapter 4.2: Environmental Compatibility
- Use of the PIRT Data
- Corrosion Corrosion Testing with Use of Compositional Analysis and Electrochemical Potential (ECP)
Monitoring
- Environmentally Assisted Cracking Slow Strain Rate Testing (SSRT)
Fracture Mechanics Based Testing -
Corrosion Fatigue (CF) and Stress Corrosion Cracking (SCC)
Environmental Creep Testing
- Metallurgical Effects / Other Stress Relaxation Cracking Phase Formation Embrittlement Thermal Cycling / Striping
- Irradiation Effects Irradiation-Induced Embrittlement Irradiation-Affected Corrosion Irradiation-Assisted Stress Corrosion Cracking (IASCC)
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11 Some Major Comments that Have Been Addressed
- Testing at 750°C added to bound maximum expected transient temperature and time
- Heat affected zone (HAZ) samples added throughout EAC and Corrosion testing and these samples are priorities for Hermes Rx testing
- Significant detail added to testing plans (e.g., numbers of samples, exposure times, etc. for corrosion tests, specific conditions for environmentally assisted cracking tests)
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No Reproduction or Distribution Without Express Written Permission of Kairos Power LLC.
12 Summary
- Alloy 316H base metal and ER16-8-2 weld filler metal have been selected as the metallic structural alloys for use in safety-significant, high temperature, Flibe-wetted component designs.
- This testing is being conducted to support the design and licensing of both the non-power test reactor (Hermes) and the commercial power generation reactor (KP-X).
- This testing is focused on structural alloys 316H and ER16-8-2 for the reactor vessel, which was determined to be the primary safety-related component of interest, as it serves to retain the Flibe coolant (a fission product barrier) around the fuel pebbles.
- Testing to support design includes work to extend the code qualification of ER16-8-2 weld metal up to a temperature to match the current qualification of Alloy 316H base metal as well as testing and analyses required to support design per the ASME Code Section III, Division 5.
- The materials testing consists of two major efforts: (1) testing in high temperature air to support ASME design and (2) testing in molten Flibe salt to account for potential environmental degradation.
NRC Evaluation of KP-TR-013-P, Metallic Material Qualification for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR), Rev. 4 John Honcharik Alex Chereskin US Nuclear Regulatory Commission January 12, 2023
Introduction 2
Kairos Power, LLC requested staff review and approval of Topical Report (TR)
KP-TR-013-P, Rev. 4, Metallic Material Qualification for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR) to address environmental affects on materials TR provides a material testing plan (including analysis and monitoring programs) for metallic structural materials used in Flibe-wetted areas for safety-related high temperature components of the KP-FHR power and non-power (test) reactors These material qualification test results will be used as a basis in future licensing actions to address potential materials reliability and environmental compatibility in KP-FHR designs 1 via design, operation, inspection and surveillance programs to manage the materials performance.
1 The term KP-FHR designs applies to both the power reactor and non-power test reactor, unless otherwise specified.
Regulatory Basis 3
Title 10 of the Code of Federal Regulations (10 CFR) Sections 50.34(a),
50.34(b), and corresponding regulations for design certification applications, combined license applications and standard design approvals The following Kairos Principal Design Criteria (PDC) are applicable to this topical report and were previously approved by the NRC staff (KP-TR-003-NP-A):
KP-FHR PDC 14, Reactor coolant boundary KP-FHR PDC 31, Fracture prevention of reactor coolant boundary
Staff Evaluation-Overview 4
The staffs review focused on the overall testing framework to conclude that there is reasonable assurance that the test plan, in part, will satisfy KP-FHR PDCs 14 and 31. Specific topics included:
Material - 316H and ER-16-8-2 stainless steel Test environment Four degradation mechanism categories:
Corrosion Environmentally assisted cracking Effects on Metallurgical properties Irradiation
Staff Evaluation-Overview Corrosion 1.
general corrosion 2.
crevice corrosion 3.
thermal aging 4.
erosion/wear 5.
cold leg occlusion
- Environmentally assisted cracking 6.
environmental creep 8.
corrosion fatigue
- Effects on Metallurgical properties 9.
stress relaxation cracking
- 10. phase formation embrittlement
- 11. thermal cycling
- Irradiation
- 12. irradiation-affected corrosion,
- 13. irradiation-assisted stress corrosion cracking,
- 14. irradiation-induced embrittlement 5
14 Degradation mechanism addressed:
Staff Evaluation-Material
- Metallic Material Used
- 316H austenitic stainless steel and associated weld material ER-16-8-2 stainless steel
- Qualification for mechanical properties of 316H and ER16-8-2 weld material are provided by ASME Code,Section III, Division 5, and endorsed in Regulatory Guide 1.87, Rev. 2 (TBA))
- Both materials are listed in Section III, Division 5 for high temperature applications (Limitation & Condition 4 - ER16 2 must be qualified to postulated accident conditions)
- TR only addresses environmental effects and does not cover the mechanical properties qualification 6
Staff Evaluation-Environment KP-FHR Environment (Operating)
- Flibe salt temperatures of 550°C-650C
- An intermediate salt coolant loop (commercial reactor)
- A Primary Heat Transport System to the air (non-power test reactor)
- Non-power test reactor lifetime of 5 years and commercial power reactor lifetime
- Near-atmospheric primary coolant pressures
- End of life irradiation of less than 0.1 displacement per atoms (dpa)
Testing environment duplicates the environment of KP-FHR designs.
Higher test temperature to develop environmental degradation rates during postulated accident scenarios Testing conducted in Nominal Flibe, (Flibe purified to minimize water and other oxidizing contaminants but not metal additions to invoke redox control)
Additional testing in Redox Controlled Flibe Limitation & Condition 3 - Time and temperature must bound RG 1.87, Table 2 7
Staff Evaluation-Environment KP-FHR Environment (Postulated accident scenarios)
- Two potential accident scenarios for the commercial power reactor (i.e.,
intermediate salt ingress and air ingress for into the Flibe salt)
- Potential accident scenario for non-power test reactor (Nominal Flibe with air as an impurity)
The NRC staff finds testing with impurities acceptable
- Power reactor - testing of 316H and ER16-8-2 in Nominal Flibe with air and intermediate salt (as an impurity) will bound accident conditions and provides a reasonable method of developing corrosion rates.
- Non-power reactor - corrosion testing of 316H and ER16-8-2 in Nominal Flibe with air (as an impurity) provides a reasonable method of developing corrosion rates for the non-power test reactor.
Limitation & Conditions 3 - Time and temperature must bound RG 1.87, Table 2 Limitation & Condition 8 - Impurity ingress testing (e.g., the concentration of contaminant) must bound safety analysis 8
Staff Evaluation-Corrosion Corrosion
- Proposed corrosion testing will be used to develop quantitative corrosion models for 316H and ER16-8-2 stainless steel in a Flibe environment.
- Proposed testing will use representative coupons of these materials in conditions described in Tables 12 and 13 of the TR.
- Tests performed under different conditions to assess the impacts of specific corrosion degradation mechanisms.
9
Staff Evaluation-Corrosion
- The NRC staff found proposed qualification testing acceptable to determine the impacts of corrosion
- These tests will determine the impact of temperature, microstructure, salt composition, geometry, erosion-corrosion, thermal aging, presence of graphite, and redox control on the corrosion rates of 316H and ER16-8-2.
- Occluded geometry effects on corrosion rates -acceptable because tests will determine whether crevice corrosion is a concern for 316H and ER16-8-2 in Flibe
- Effect of temperature on corrosion rates - acceptable since tests are evaluated over a range of temperatures consistent with the operating temperatures of the KP-FHR designs including bounding postulated accident conditions 10
Staff Evaluation-Corrosion
- The NRC staff found proposed qualification testing acceptable to determine the impacts of corrosion (contd)
- Tests are consistent with the expected corrosion mechanisms for 316H and ER16-8-2 in a molten salt environment and a portion of the tests will be conducted with flowing Flibe, which is necessary to simulate the flowing salt in a reactor.
- Tests to determine the effect of erosion-corrosion acceptable because the tests will utilize graphite particulate to determine the effect of these particles on corrosion rates
- Impact of microstructure effects (heat affected zone, thermal aging and cold work) will be tested and evaluated.
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Staff Evaluation-Environmentally Assisted Cracking 12 Stress Corrosion Cracking and Corrosion Fatigue The NRC staff found the proposed testing plan provides reasonable assurance in determining the crack growth rates for fatigue and stress corrosion cracking relative to the environment in the KP-FHR designs:
In-situ mechanical testing systems will be used to conduct slow strain rate testing (SSRT) for corrosion fatigue and stress corrosion cracking in both normal and accident conditions The samples to be tested are acceptable because they are representative of the material and weldments to be used in the KP-FHR designs.
Staff Evaluation-Environmentally Assisted Cracking 13 Environmental Creep The NRC staff found the proposed testing plan acceptable:
Creep testing in both Normal Flibe and air using welded base metal samples to include the base metal, weld metal and heat affected zone Creep tests in Nominal Flibe to determine if the Flibe contributes additional degradation beyond those determine from the creep tests performed in air.
If the testing determines Flibe has an additional effect on degradation, additional testing would be required to quantify any increase in degradation contributing to Flibe
Staff Evaluation-Metallurgical Effects 14 Stress Relaxation Cracking The NRC staff found the proposed testing plan acceptable:
Current data and experience show that ER16-8-2 weld metals are not susceptible to stress relaxation cracking, while the heat affected zone of 316H base metal with high triaxial stresses is susceptible Testing in air is valid for 316H in Flibe for the KP-FHR designs since triaxial stresses are the major contributor to stress relaxation cracking and not environment Comparing susceptibility of 316H to that of 347 would allow a determination of the bounding triaxial stresses Using test results to conduct future analysis and design refinements of the KP-FHR designs, such as weld designs and specific weld processes and parameters to reduce triaxial stresses thereby minimizing stress relaxation cracking
Staff Evaluation-Metallurgical Effects 15 Phase Formation Embrittlement The NRC staff found the proposed testing plan acceptable:
Phase formation embrittlement may occur when 316H and ER16-8-2 picks up an element during its exposure to Flibe and forms a deleterious second phase Testing will determine whether this degradation mechanism occurs for the KP-FHR designs Limitation & Condition 11 - If intermetallic formation occurs, an applicant will need to perform testing to quantify the effects on the mechanical properties of 316H and associated weld material ER16 2
Staff Evaluation-Metallurgical Effects 16 Thermal Cycling/Striping The NRC staff found the proposed testing plan acceptable:
Degradation of 316H and ER16-8-2 by large thermal transients could lead to high stresses resulting in thermal fatigue degradation Thermal cycling will be addressed by conducting analysis to refine the design and operation of the KP-FHR designs to mitigate large thermal gradients.
Limitation & Condition 12 - Assess thermal cycling/striping in future licensing submittals
Staff Evaluation-Irradiation Effects 17 Irradiation-Induced Embrittlement The NRC staff found the proposed testing plan acceptable:
At low strain rates, data shows irradiation-induced embrittlement can affect tensile strength, ductility and creep life due to the generation of helium.
Existing data will be used to develop degradation factors Conduct irradiation tests on ER16-8-2, 316H, and the associated heat affected zone of 316H to quantify margins at irradiation levels for the non-power test reactor and the commercial power reactor Limitation & Condition 13 - Test environment shall bound the KP-FHR designs, including the expected irradiation damage (dpa) and helium content
Staff Evaluation-Irradiation Effects 18 Irradiation-Affected Corrosion The NRC staff found the proposed testing plan acceptable:
Existing data shows that irradiation may increase general corrosion rates Materials surveillance system program for the non-power test reactor and the commercial power reactor systems to monitor irradiation-affected corrosion.
Inspection and monitoring program that will assess the wall thickness of the reactor vessel will also be implemented Existing data will be used to develop degradation factors and monitored by surveillance Limitation & Condition 14 - A materials surveillance program and an inspection and monitoring program must be implemented
Staff Evaluation-Irradiation Effects 19 Irradiation-Assisted Stress Corrosion Cracking The NRC staff found the proposed testing plan acceptable:
IASCC is not expected to be a degradation mechanism in the KP-FHR design due to the low irradiation level (<0.1 dpa)
Stress corrosion test program specified in Section 4.2.4 will determine if stress corrosion cracking is a credible degradation mechanism for the environment of the KP-FHR designs Materials surveillance system program for the non-power test reactor and commercial power reactor systems to monitor IASCC.
Limitation & Condition 14 - A materials surveillance program and an inspection and monitoring program must be implemented
20 Conclusions
- With limitations and conditions noted, material testing plan (including analyses, surveillance and monitoring) for 316H and ER16-8-2 will provide necessary information to address the materials reliability and compatibility in the environment of the KP-FHR designs:
Testing duplicates the operating environment (normal and accident) that the material will experience in the KP-FHR designs Material test samples are representative of actual weldments Analyses for stress relaxation cracking and thermal cycling Surveillance for irradiation effects (corrosion and stress corrosion cracking)
Results of planned tests and analyses will be used for future license applications to ensure extremely low probability of abnormal leakage, rapidly propagating failure, or gross rupture, which partially satisfies PDC 14 and PDC 31
21 Questions?
Back-Up Slides 22 LIMITATIONS AND CONDITIONS (Redacted)
An applicant may reference the TR only if the applicant demonstrates compliance with the following limitations and conditions:
1.
(Section 1.0) As stated by Kairos in the TR, NRC staff review and approval of only Section 4 of the TR was requested. Therefore, KP-FHR designs referencing this TR may only use this TR for purposes related to the information on 316H and ER16-8-2 material found in Section 4 of the TR, subject to the specific limitations and conditions found in the NRC staff SE below. All other information related to the 316H and ER16-8-2 material will be evaluated in separate documents and licensing actions.
2.
(Sections 1.1.3.2 and 5.1) The environmental effects qualification testing for the KP-FHR designs in this TR can only be used for other components with environments that are bounded by the environment the reactor vessel would experience and are used in this TR. For example, other components that would have Flibe on one side of the metallic material and another salt on the other side of the metallic material, or higher irradiation levels than those specified in the TR, etc. would not be bounded by this TR.
3.
If the time and temperature for both normal operations and postulated accident conditions change for the KP-FHR designs, they must still be bounded by the NRC staff-endorsed ranges found in Table 2 of Regulatory Guide 1.87 for 316H, or an adequate justification must be provided for NRC staff review and approval for why the values outside of the endorsed ranges are acceptable.
Back-Up Slides LIMITATIONS AND CONDITIONS (Redacted; Cont.)
4.
(Section 4.2.1) ER16-8-2 material must be qualified to a temperature of ((
)) in accordance with the requirements of ASME Code,Section III, Division 5, and for a time that bounds the postulated accident conditions and be approved by the NRC staff.
5.
(Section 1.1) Because there is information that has not yet been developed and/or reviewed as part of this TR, KP-FHR designs referencing this TR must provide information that completely and accurately describes the design of the reactor coolant boundary (and associated systems) and any associated functions it is credited to perform for NRC staff review and approval. As stated in the TR, if key design features of the KP FHR designs change, or if new or revised regulations are issued that impact descriptions and conclusions in this TR, these changes would be reconciled and addressed in future license application submittals. Due to the potential for design changes and new or revised regulations, KP FHR designs referencing this TR must demonstrate that all regulatory and safety requirements related to the characteristics of the metallic materials are met when considering the final design of the KP FHR.
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Back-Up Slides LIMITATIONS AND CONDITIONS (Redacted; Cont.)
6.
(Section 4.1) As presented in the TR, there are key design parameters without which the proposed reactor coolant boundary design and associated properties may not be supported. Therefore, KP-FHR designs referencing this TR must have the following:
Flibe Salt temperatures of 550°C-650C An intermediate salt coolant loop for the commercial reactor A Primary Heat Transport System that rejects heat to the air in lieu of an intermediate coolant loop for the non-power test reactor Non-power test reactor lifetime of a maximum of 5 years (1 year commissioning + 4 years operation) and commercial power reactor lifetime of a maximum of (( ))
Near-atmospheric primary coolant pressures End of life irradiation of less than 0.1 dpa These key design parameters of the KP FHR designs, if changed, could necessitate the modification of, or addition to, the testing program.
7.
(Tables 12, 13, 14, 15 and 16) If the postulated accident conditions ((
)), the test temperature and time for the associated material testing in Section 4 of the TR must be increased to (( )).
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Back-Up Slides LIMITATIONS AND CONDITIONS (Redacted; Cont.)
8.
(Section 4.2.3.3 and Table 13) The impurity effects testing on 316H and ER16-8-2 must include the potential loss of Flibe chemistry control from both air ingress and intermediate salt loop ingress based on the safety analysis reports. An applicant referencing this TR must demonstrate that any potential impurity ingress (including postulated accidents) in the KP-FHR designs is bound by the testing performed as part of this TR.
9.
(Section 4.2.3.2, Tables 13 and 14) An applicant referencing this TR must demonstrate that the Nominal Flibe salt composition used in the KP-FHR designs is consistent with the Nominal Flibe salt composition used in these tests including initial impurities in the salt.
10.
Section 4.2.3.2, Tables 13 and 14) An applicant referencing this topical report must demonstrate that the salt compositions (with reducing agent additions and impurities from postulated accident scenarios) tested in this program bound any potential salt compositions for the KP-FHR reactor designs.
11.
(Section 4.2.5) In order to address phase formation embrittlement for the KP-FHR designs an applicant must show that testing bounds potential design conditions
)) and that if a secondary phase is detected during testing, the effects on mechanical properties of 316H and ER16-8-2 must be quantified via testing and approved by the NRC staff.
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Back-Up Slides LIMITATIONS AND CONDITIONS (Redacted; Cont.)
12.
(Section 4.2.5 and Table 11) The applicant will assess thermal cycling/striping in future licensing submittals by minimizing the thermal gradients via appropriate design and operating conditions of the KP-FHR designs based on analysis.
13.
(Section 4.2.6.1) Testing for irradiation-induced embrittlement of ER16-8-2, 316H, and the associated heat affected zone of 316H must be performed that bounds the environment representative of the KP-FHR designs, including the expected irradiation damage (dpa) and helium content. The program describing this testing must be submitted in future license applications for NRC staff review and approval to verify this testing program is sufficient to address irradiation-induced embrittlement of the reactor vessel.
14.
(Sections 4.2.6.2 and 4.2.6.3) As described in Sections 4.6.2.2 and 4.2.6.3 of the TR, a materials surveillance program and an inspection and monitoring program must be implemented for all non-power test reactors and commercial power reactors using KP-FHR designs to assess and monitor both irradiation-affected corrosion rates and irradiation-affected stress corrosion cracking rates of 316H and ER16-8-2 in the environment of KP-FHR designs. The materials surveillance program and the inspection and monitoring program must be submitted in future license applications for NRC staff review and approval to verify these programs are sufficient to address both irradiation-affected corrosion and irradiation-affected stress corrosion cracking of the reactor vessel.
15.
(Section 1.0) Material testing for the commercial power reactor must be conducted under quality assurance program that meets the requirements of 10 CFR Part 50 Appendix B to confirm the quality of the data obtained during the material testing that will be used for the commercial power reactor.
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Back-Up Slides 27