ML21257A324

Transcript of the Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Meeting, July 20, 2021, Pages 1-109

ML21257A324
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Issue date:	07/20/2021
From:	Advisory Committee on Reactor Safeguards
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Howard, K, ACRS
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NRC-1595
Download: ML21257A324 (163)

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Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION

Title:

Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Docket Number:

(n/a)

Location:

teleconference Date:

Tuesday, July 20, 2021 Work Order No.:

NRC-1595 Pages 1-109 NEAL R. GROSS AND CO., INC.

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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 FUTURE PLANT DESIGNS SUBCOMMITTEE 7

+ + + + +

8 TUESDAY 9

JULY 20, 2021 10

+ + + + +

11 The Subcommittee met via Videoconference, 12 at 9:30 a.m. EDT, Peter Riccardella, Chair, presiding.

13 14 COMMITTEE MEMBERS:

15 PETER RICCARDELLA, Chair 16 RONALD G. BALLINGER, Member 17 VICKI M. BIER, Member 18 CHARLES H. BROWN, JR. Member 19 GREGORY H. HALNON, Member 20 WALTER L. KIRCHNER, Member 21 JOSE MARCH-LEUBA, Member 22 DAVID A. PETTI, Member 23 JOY L. REMPE, Member 24 MATTHEW W. SUNSERI, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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2 ACRS CONSULTANT:

1 STEPHEN SCHULTZ 2

3 DESIGNATED FEDERAL OFFICIAL:

4 KENT HOWARD 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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3 C-O-N-T-E-N-T-S 1

PAGE 2

Opening Remarks by Peter Riccardella, ACRS 4

3 Staff Remarks by Louise Lund, RES........

7 4

NRC Staff Presentation 5

Overview of ASME Code Section III, 6

Division 5

................... 10 7

NRC Staff Presentation 8

Staff Efforts on Potential Endorsement 9

of ASME Code Section III, Division 5...... 67 10 Adjourn....................

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4 P R O C E E D I N G S 1

9:30 a.m.

2 CHAIR RICCARDELLA: This is a meeting of 3

the Future Plant Designs Committee. The meeting will 4

now come to order. I am Pete Riccardella, Chairman of 5

this meeting. ACRS members in attendance are Ron 6

Ballinger, Dave Petti, Joy Rempe, Walt Kirchner, Vicki 7

Bier, Matt Sunseri, Greg Halnon, and Charles Brown.

8 Is our consultant, Steve Schultz -- are you on the 9

meeting?

10 (No response.)

11 CHAIR RICCARDELLA: Okay. Steve was 12 expected to join. He might be on soon.

13 DR. SCHULTZ: I'm here, Pete.

14 CHAIR RICCARDELLA: Okay. And our 15 consultant, Steve Schultz, is also in attendance.

16 Kent Howard of the ACRS staff is the Designated 17 Federal Official for this meeting.

18 The purpose of today's meeting is an 19 information briefing from the NRC staff on potential 20 endorsement of ASME Section III, Division 5, High 21 Temperature Reactors. The subcommittee will gather 22 information, analyze relevant issues and facts, and 23 formulate proposed positions and actions as 24 appropriate.

However, at the subcommittee's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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5 direction, any matter will be considered for 1

presentation at the full committee if necessary as the 2

members see fit. The ACRS was established by statute 3

and is governed by the Federal Advisory Committee Act, 4

FACA.

5 The NRC implemented FACA in accordance 6

with regulations found in Title 10 of the Code of 7

Federal Regulations, Part 7. The committee can only 8

speak to its published letter reports. We hold 9

meetings to gather information and perform preparatory 10 work that will support our deliberations at a full 11 committee meeting, if necessary.

12 The rules for participating in all ACRS 13 meetings, including today's, were announced previously 14 in the Federal Register. The ACRS section of the U.S.

15 NRC public website provides our charter, bylaws, 16 agendas, letter reports, and full transcripts of all 17 full and subcommittee meetings, including slides 18 presented there. The meeting notice and agenda for 19 this meeting were posted there.

20 Members of the public who desire to 21 provide written or oral input to the subcommittee may 22 do so and should contact a designated federal official 23 five days prior to the meeting as practicable.

24 Today's meeting is open to the public attendance. And 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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6 there will be time set aside during the meeting for 1

spontaneous comments from members of the public 2

attending or listening to our meetings.

3 Due to the COVID pandemic, today's meeting 4

is being held over Microsoft Teams for ACRS, NRC, and 5

members of the public. There is also a telephone 6

bridgeline allowing participation of the public over 7

the phone. This public bridgeline is controlled by 8

the ACRS staff and should not be muted by anyone other 9

than the designated ACRS staff members.

10 A transcript of today's meeting is being 11 kept. Therefore, we will request that meeting 12 participants on the bridgeline identify themselves 13 when they are asked to speak and to speak with 14 sufficient clarity and volume so that they can readily 15 be heard. At this time, I ask that attendees on the 16 Teams and bridgeline mute their phones to minimize the 17 disruption and to unmute your individual devices only 18 when speaking.

19 We will now proceed with the meeting. I 20 call on Louise Lund, Division Director of the Division 21 of Engineering, Office of Nuclear Regulatory Research, 22 to make introductory remarks. Louise, are you there?

23 MS. LUND: Yes, thank you. Thank you, Dr.

24 Riccardella, and good morning to the ACRS members and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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7 all others here for this meeting. I hope everybody 1

can hear me. Can I be heard?

2 CHAIR RICCARDELLA: You're fine, Louise.

3 MS. LUND: Great, wonderful. So I'm 4

Louise Lund, Director of the Division of Engineering, 5

Office of Nuclear Regulatory Research. I also serve 6

as the Agency standards executive for the codes and 7

standards

program, coordinating the Agency 8

participation on various standard development 9

organization committees, and assuring Agency goals and 10 activities relative to staff participation and 11 development and use of consensus standards.

12 On behalf of the staff, we are very 13 pleased to have the opportunity to present on the 14 review and potential endorsement of the ASME Boiler 15 and Pressure Vessel Code,Section III, Division 5, 16 high temperature reactors. As you know, the NRC is 17 executing its vision to become a modern risk informed 18 regulatory by developing approaches to streamline and 19 optimize reviews to enable the deployment of advanced 20 reactor technologies. As part of the vision, the NRC 21 developed implementation action plans for various 22 strategic areas.

23 Consistent with its implementation action 24 plans, NRC has been working proactively towards 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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8 enhancing its non-LWR technical readiness and 1

optimizing regulatory readiness. Today's presentation 2

will be part of Strategy 4 which aims at facilitating 3

development of industry codes and standards. To 4

further that objective, the staff developed a prudent 5

and balanced approach to ensure efficient completion 6

of the endorsement project.

7 The approach involved building staff 8

knowledge through training and collaborative 9

activities, active participation in the ASME Section 10 III working groups, engaging contractors to perform 11 reviews and provide recommendations, and performing 12 independent assessment of the code rules and 13 procedures and contractor recommendations.

14 Recognizing that the technical expertise on high 15 temperature materials and components for advanced non-16 light water reactors was largely confined to a small 17 group of people who were involved in the code 18 development. But staff engaged these experts to seek 19 clarification on staff's assessment and contractors' 20 recommendations where applicable.

21 With such a comprehensive approach, the 22 staff has pursued a holistic and balanced endorsement 23 of the ASME Section III, Division 5 code. This review 24 represents a major collaborative and successful 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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9 undertaking by the staff across multiple divisions in 1

both NRR and research. And we anticipate that the 2

endorsement of the ASME high temperature provisions 3

for use by a prospective non-LWR vendors will improve 4

the efficiency and effectiveness at the NRC's review 5

process.

6 Thank you again for the opportunity to 7

present. And we look forward to our discussions this 8

morning. Now Jeff Poehler of my staff will provide an 9

overview of the ASME Code Section III, Division 5.

10 Jeff?

11 MR. POEHLER: Good morning, everyone. Can 12 you hear me well?

13 MS. LUND: Yes.

14 CHAIR RICCARDELLA: I hear you fine.

15 (Simultaneous speaking.)

16 MR. POEHLER: Yeah, I'll turn my camera 17 off in a minute because I know you guys probably don't 18 want to look at me too much but just so you know who 19 I am. Yeah, so I'm going to be presenting an overview 20 of Section III, Division 5, trying to give a high 21 level overview and just give you a flavor of what it's 22 about. I find that the Division 5 code is kind of 23 hard to get your hands around.

24 There's a lot to it, even for people that 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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10 are familiar with codes and standards. I've been 1

immersed in it for about a year and a half, and I'm 2

still -- frankly, still learning. So I am going to 3

call on project team members if needed for questions, 4

and also we have some experts from the national labs.

5 Sam Sham and Will Windes from Idaho 6

National Laboratory are in the meeting. So I may 7

throw some questions to them. But anyway, I would 8

like to also thank the project team for all their help 9

preparing these presentations.

10 And also this is the first presentation.

11 The second presentation, we'll focus on the review 12 process and potential exceptions and limitations to 13 our review. So next slide, please. Okay. So I'm 14 going to discuss the scope of Division 5.

15 So the scope of Division 5 governs 16 construction of vessels, piping, pumps, valves, 17 supports, core support structures, and nonmetallic 18 core components for use in high temperature reactor 19 systems and their supporting systems. And term, 20 construction, here includes

material, design, 21 fabrication, installation, examination, testing, over-22 pressure protection, inspection,

stamping, and 23 certification, so basically the same areas covered by 24 the low temperature construction code in Section III, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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11 Division 1. And high temperature reactors includes a 1

wide variety of designs including gas-cooled reactors, 2

liquid metal cooled reactors, and molten salt 3

reactors.

4 Division is inclusive of all these 5

technologies, meaning it's not specific to any of the 6

particular reactor technologies. Let's go to the next 7

slide, please. And this slide just kind of shows the 8

spectrum of some of the advanced reactor designs that 9

are being developed by the industry which span from 10 fast reactors to gas reactors, heat pipe reactors.

11 You have molten salt reactors, and those 12 can be either molten salt cooled and also molten salt 13 fueled. And you have also -- you have fast and 14 thermal reactors in this spectrum. So it's a lot of 15 different types. Let's go to the next slide.

16 So Division 5 is a component code, and 17 this is basically high level how it's organized.

18 Class A is the highest safety class. The classes --

19 Class A is analogous to Class 1 in Division 1, and 20 Class B is analogous to Class 2 in Division 1.

21 You also have Class SM for metallic core 22 supports. And then you have Class SN for non-metallic 23 core supports which at this point essential means 24 graphite core support structures. And Division 5 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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12 recognizes different levels of importance associated 1

with a function of each component as related to the 2

safe operation of the advanced reactor plant.

3 So these code classes allow a choice of 4

rules that provide a

reasonable assurance of 5

structural integrity and quality in line with the 6

relative importance assigned to the individual 7

components of the advanced reactor plant. Next slide, 8

please. So this slide covers some of the things that 9

Division 5 does not address, and those include 10 corrosion, irradiation, mass transfer phenomena which 11 would include things erosion and flow accelerated 12 corrosion, radiation

effects, other material 13 instabilities which could be metallurgical phenomena.

14 It also doesn't cover continued functional 15 performance of deformation sensitive structures such 16 as valves and pumps. And what that means to me is it 17 doesn't address whether the moving parts actually 18 move. But let's go to the next slide. Just a little 19 history now.

20 So there's a lot of history with the 21 development of the high temperature rules which it's 22 too much to go through in detail with the time we 23 have. But the design rules do stretch all the way 24 back to the 1960s with Code Case 1331. But really 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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13 what I want to focus on the 1590 series code cases 1

which were developed in the early '70s.

2 Those were reviewed by the NRC and 3

endorsed in Regulatory Guide 1.87, Revision 1 which 4

came out in June 1975. And that endorsed those code 5

cases with conditions. And then later, the code case 6

series, 1592 through 96, were converted into Code Case 7

N-47, and that later formed the basis for Section III, 8

Division 1, Subsection NH which cover high temperature 9

components. NRC never reviewed N-47.

10 And then Division 5 was first published in 11 2011, and it combined Subsection NH and some other 12 high-temperature code cases and also the rules for 13 graphite core components which were completely new.

14 They had never been in a code case before. Next 15 slide, please.

16 So I call this slide the magic decoder 17 ring for the organization of Division 5. I'm not 18 going to go through it in detail. But I will point 19 that for each subsection on metallic components, there 20 are subparts for low temperature and elevated 21 temperature service.

22 So Subpart A would be low temperature 23 service. Subpart B would be elevated temperature 24 service. And that holds for Class A metallic 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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14 components, Class B metallic pressure boundary 1

components, and also for core support -- metallic core 2

support.

3 But the general requirements depart from 4

that pattern with Subpart A being general requirements 5

for metallic materials, Subpart B being general 6

requirements for graphite and composite materials.

7 And then when you get to graphite which is Subsection 8

HH, you have Subpart -- or actually Subsection HH is 9

not a metallic core component. So Subpart A of that 10 would be graphite material. Subpart B is composite 11 materials. Next slide.

12 So on this slide, I'm going to attempt to 13 explain the temperature boundaries for low and high 14 temperature reactor components under Section III, 15 Division 5. This graph kind of explains the theory of 16 when the high temperature rules are applied. So if 17 you look at the table at bottom of the slide, it gives 18 the temperatures.

19 And those are the temperature boundaries 20 below which you can use the low temperature rules but 21 above which you have to use the elevated temperature 22 rules. Then the figure at the top here shows the 23 different temperature regimes versus time. You see 24 below a certain temperature, that's the temperatures 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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15 corresponding to this table. The blue region here, 1

you have no creep effects at all. So you can use the 2

low temperature rules.

3 Above that line in the red and yellow 4

regions, you do have creep going on in the yellow 5

region which is at lower times and lower temperatures.

6 You have creep going on but it doesn't affect cyclic 7

life, whereas in the red region at longer times and 8

higher temperatures creep does affect cyclic life. So 9

you have a creep fatigue interaction.

10 MEMBER BALLINGER: This is Ron Ballinger.

11 Where is 617 on this table?

12 MR. POEHLER: So 617 is addressed by a 13 couple of code cases. So it's not actually in 14 Division 5 itself. So I would have to look -- I could 15 look up -- I would have to look up the maximum -- the 16 temperature boundary for 617. But there is one.

17 There is both a low temperature code case and a high 18 temperature --

19 (Simultaneous speaking.)

20 CHAIR RICCARDELLA: Just for information, 21 what does 617 mean?

22 MEMBER BALLINGER: It's --

23 MR. POEHLER: Go ahead, Ron.

24 MEMBER BALLINGER: No, go ahead. Go 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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16 ahead.

1 MR. POEHLER: Well, it's a nickel-based 2

alloy that has very good high temperature strength, 3

that has been qualified for use in high temperature 4

reactors through a couple of code cases. Actually, 5

there's one case for lower temperature use and then 6

one for higher temperature use. And --

7 (Simultaneous speaking.)

8 MEMBER BALLINGER: It's not a -- it's a 9

nickel, chrome, iron, cobalt alloy.

10 MR. POEHLER: Oh, okay.

11 MEMBER BALLINGER: And the code case for 12 that -- the high temperature code case took -- oh, 13 man. It took a very, very long time to get done.

14 CHAIR RICCARDELLA: And any idea what that 15 cutoff temperature is, the Tmax is for that alloy?

16 MEMBER BALLINGER: It's got to be above 17 800 Fahrenheit for sure.

18 MR. POEHLER: We can get that for you.

19 It's --

20 MEMBER BALLINGER: Will Windes -- Will 21 would probably know. And so probably Will would know.

22 But I don't see Richard Wright on this list either.

23 He was the guy that was in charge of --

24 DR. SHAM: Ron, this is Sam Sham. So the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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17 code boundary between low temperature and high 1

temperature is just like the other or take the 2

stainless steel that is 800 degree Fahrenheit.

3 CHAIR RICCARDELLA: Okay 4

DR. SHAM: And the maximum use temperature 5

is 1750 degrees Fahrenheit.

6 CHAIR RICCARDELLA: Okay.

7 DR. SHAM: So it's around 154 degrees 8

Celsius.

9 CHAIR RICCARDELLA: Okay. Thank you.

10 MR. POEHLER: Thanks, Sam.

11 MEMBER BROWN: Pete, can I ask a question 12 on this? This is Charlie.

13 CHAIR RICCARDELLA: Sure. Go ahead.

14 MEMBER BROWN: Yeah, Ron popped up and 15 said this new alloy is what, nickel, chromium, iron, 16 cobalt?

17 CHAIR RICCARDELLA: Yes.

18 MEMBER BROWN: Is there a reason we're 19 reintroducing cobalt into a radiated material such 20 that we -- in my old program, we tried to get cobalt 21 out of everything.

22 MEMBER BALLINGER: Yeah, this 617 is not 23 used -- would not be used in a neutron environment.

24 MEMBER BROWN: Oh, okay. All right. That 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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18 wasn't clear to me. Pardon my question then. Thank 1

you.

2 MR. POEHLER: Okay.

3 MEMBER BROWN: That's it.

4 MR. POEHLER: All right. Next slide, 5

please. Okay. So this slide talks about the 6

materials that are allowed for Class A metallic 7

materials in Division 5. There's a limited set of 8

materials. There's only six materials and not 9

included Alloy 617.

10 But those are Type 304 stainless steel, 11 316 stainless steel, Alloy 800H, 2.25Cr-1Mo, and 9Cr-12 1Mo-V which is commonly known as Grade 91. And just 13 a note about the two stainless steels, Division 5 14 specifies the minimum carbon content of 0.04 weight 15 percent for those alloys to give them better high 16 temperature properties. And they are commonly called 17 Type 304H and Type 316H for that reason.

18 But that designation is not used in 19 Section III, Division 5. But you will hear 304H and 20 316H. And the design parameters for the alloys are 21 mostly in Division 5. But some of them are also 22 contained in Section II and listed at the bottom of 23 this slide. Next slide, please.

24 MEMBER BALLINGER: This is Ron again.

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19 Sam, you said that the limit is -- for 617, the 1

boundary is still 800 Fahrenheit?

2 DR. SHAM: Yes, going from Division 1 3

rules of the light water reactor. The design rules 4

for high temperature is 800 Fahrenheit --

5 MEMBER BALLINGER: So that just --

6 DR. SHAM: -- maximum.

7 MEMBER BALLINGER: -- means the allowable 8

stresses must be higher then, right?

9 DR. SHAM: The allowable stresses in the 10 creep regime is higher.

11 MEMBER BALLINGER: Okay.

12 MR. POEHLER: Okay. Anymore questions on 13 that slide? No? Next slide, please. Oh, you're on 14

-- no, you're on the right slide. Never mind.

15 So this slide kind of breaks down all the 16 different failure modes addressed by Section III, 17 Division 5, and specifically for the Class A materials 18 which is HBB. So it I didn't say it before, this 19 presentation is going to focus heavily on the Class A 20 metallic materials and also on graphite.

21 We are going to touch on Class B metallic 22 materials and core supports but to a limited extent.

23 So the majority of this is going to be about Class A 24 metallics. And that's what this slide is talking 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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20 about, the failure modes that are covered and also the 1

type of -- that they're considered, what analysis --

2 or what areas of the code prevent those failure codes, 3

where those are located, and the analysis method. So 4

the two major types of failure modes are load-5 controlled which are just in HBB-3000 and deformation-6 controlled which are addressed in non-mandatory 7

appendix HBB-T.

8 CHAIR RICCARDELLA: So Jeff, these are 9

analogous to what we used to call primary and 10 secondary stresses in Section III, Div. 1?

11 MR. POEHLER: Right. The HPV-3000 rules 12 are going to consider primary stresses.

13 CHAIR RICCARDELLA: Okay.

14 MR. POEHLER: So -- and then what we would 15 consider secondary would be addressed more in the non-16 mandatory appendix HBB-T.

17 CHAIR RICCARDELLA: Understand. Thank 18 you.

19 MR. POEHLER: So, load-controlled are 20 those quantities evaluated against the allowable 21 stresses for primary loads. And those are all 22 evaluated using elastic analysis methods. Evaluation 23 of deformation-controlled quantities is called out in 24 HBB-3250, and that allows the provisions of non-25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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21 mandatory appendix HBB-T to be used. But it also 1

allows alternative methods --

2 CHAIR RICCARDELLA: I understand.

3 MR. POEHLER: -- which is why it's not a 4

non-mandatory appendix. But these quantities include 5

strains and deformations, ratcheting and creep 6

fatigue. Buckling is also addressed in HBB-T, and 7

that can be either load-controlled, strain-controlled, 8

or a combination of both. And as I mentioned, in HBB-9 3000 rules, only elastic analysis allowed whereas in 10 HBB-T, it allows either elastic analysis, inelastic 11 analysis, and also elastic, perfectly plastic analysis 12 which is allowed through the two code cases.

13 CHAIR RICCARDELLA: Yeah.

14 MR. POEHLER: So okay. Next slide, 15 please. So this slide attempts to highlight the 16 general characteristics of the HBB primary load design 17 on the left and then the evaluation of design loads 18 versus loads on the right. So generally, HBB primary 19 load design has the following characteristics. It's 20 based on elastic analysis, load-controlled, uses 21 stress classification and linearization, includes 22 design and service level load checks.

23 It accounts for thermal aging effects with 24 factors on yield and ultimate strength. And for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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22 welds, there is a strength reduction factor applied.

1 And then on the right-hand graphic here with respect 2

to design of service loads, so design loads are 3

evaluated in a single temperature, pressure, and set 4

of forces. They're time independent, and they use the 5

allowable stress, S sub 0.

6 The procedures are very similar to those 7

to the Section III -- I'm sorry,Section I and Section 8

XIII of the ASME Boiler and Pressure Vessel Code. The 9

service loads evaluation accounts for the time history 10 of loading and are compared to time dependent 11 allowable stresses. And that methodology is unique to 12 Division 5. But I'm going to talk about that more on 13 some subsequent slides.

14 CHAIR RICCARDELLA: For the surface loads, 15 do we have different services levels as we did --

16 MR. POEHLER: Yes, it addresses Service 17 Level A and B, C and D.

18 CHAIR RICCARDELLA: Okay. Thank you.

19 MR. POEHLER: Yeah, thanks. Next slide.

20 Now I'm going to get into the allowable stresses a 21 bit. So you have both time dependent and time 22 independent level stresses. S sub 0 is the allowable 23 stress for design loadings.

24 The service level loading allowable 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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23 stresses include S sub m which is a time independent 1

allowable stress, S sub t which is a time dependent 2

level stress, and then S sub mt, the allowable limit 3

for general primary membrane stress for Surface Level 4

A and B. And that is determined by the lower of S sub 5

m and S sub t. And then also you have S sub r, the 6

expected minimum stress-to-rupture. That's used in 7

the Level D limits and then also in the deformation-8 controlled analyses of HBB-T. Or I guess I should say 9

used directly in some of those analysis.

10 CHAIR RICCARDELLA: So is the S sub 0 --

11 are there values above the cutoff, the 700 and 800 12 degree cutoff temperatures?

13 MR. POEHLER: Yes, sir. And I'm going to 14 discuss that a little more on the next --

15 CHAIR RICCARDELLA: Okay. All right.

16 Thank you.

17 MR. POEHLER: So next slide, please.

18 Yeah, so the basis for allowable stresses, so both S 19 sub 0 and S sub m are essentially based on Section II, 20 Part D values, either directly or extended using the 21 same methodology for higher temperatures.

22 CHAIR RICCARDELLA: Okay.

23 MR. POEHLER: And so the S criteria in 24 Section II-D may be controlled by the 100,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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24 rupture stress or stress to produce a creep rate of 1

0.01 percent in 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. So it takes into account 2

creep to some extent. So I guess what I should've 3

pointed out that S sub 0 is equal to the higher of the 4

S values from Section II-D, Subpart 1, Table 1A, or 5

the 300,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> S sub mt value which generally would 6

only be controlling in rare cases.

7 And then the S sub m is basically from 8

Section II-D, Table 2A, the S sub m values in that 9

table at the lower temperatures and then it's extended 10 to higher temperatures using the same criteria in 11 Division 5. And S sub t as I mentioned is the lower 12 of the S sub m or time independent and the S sub t 13 time dependent allowable stress. I'm going to talk 14 about how S sub t is determined on the next slide.

15 CHAIR RICCARDELLA: Okay.

16 MR. POEHLER: So next slide, please. So 17 as I said, S sub t is determined by the lowest of 18 three different quantities. Those are 100 percent of 19 the average stress required to obtain a total elastic 20 primary -- plastic primary and secondary creep strain 21 of 1 percent, or 80 percent of the minimum stress 22 causes initiation of tertiary creep, or 67 percent of 23 the minimum stress to cause rupture or S sub r.

24 And the determination of S sub t is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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25 inherently conservation because of the 80 percent and 1

67 percent factors applied to tertiary creep 2

initiation and stress-to-rupture. Also, it has been 3

noted that of those three criteria, only the 67 4

percent of rupture stress criteria is directly related 5

to component failure. The other two criteria are sort 6

of different, semi-arbitrary points from the creep 7

curves.

8 CHAIR RICCARDELLA: Yeah.

9 MR. POEHLER: So it is conservative.

10 (Simultaneous speaking.)

11 CHAIR RICCARDELLA: -- have much time 12 until you -- right? I mean, that's when the curve 13 turns up and you have not that much time until 14 rupture, right?

15 MR.

POEHLER:

Right, yeah.

It's 16 theoretically. But some materials don't exhibit 17 classical creep behavior. And it also can be 18 difficult to determine the onset of tertiary creep in 19 materials that don't have classic tertiary creep 20 behavior. I'm going to talk about that a little more 21 later.

22 CHAIR RICCARDELLA: Okay. Thank you.

23 MR. POEHLER: Next slide. And just a few 24 of the other stresses and material properties, you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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26 have yield strength and ultimate strength which are 1

self-explanatory. They are extended to the higher 2

temperatures. You have -- the R factors are weld 3

strength reduction factors to account for the reduced 4

strength of welds compared to the corresponding base 5

metal.

6 You also have tensile and yield strength 7

reduction factors apply to some materials. And those 8

account for thermal aging those materials. You also 9

have isochronous stress-strain curves which provide 10 stress versus strain curves for various times up to 11 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. And those curves are derived from 12 creep data. They're used in the analysis of some of 13 the deformation-controlled quantities and non-14 mandatory appendix HBB-T.

15 CHAIR RICCARDELLA: So 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is 16 about 35 years --

17 MR. POEHLER: Yeah.

18 CHAIR RICCARDELLA: -- for picking that 19 time?

20 MR. POEHLER: I'm not sure what the reason 21 was. I might throw that question to Sam Sham.

22 DR. SHAM: Oh, yes. At the time that we 23 sort of look at sort of the design of 40 years, 24 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> with availabilities of roughly close to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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27 that. And so currently, ASME is looking into 1

extending the allowable stresses to support a longer 2

design lifetime by 60 years.

3 CHAIR RICCARDELLA: Okay, okay. So it's 4

basically the 40-year lifetime at some availability 5

level or something.

6 DR. SHAM: Yeah, something like that.

7 CHAIR RICCARDELLA: I got it. Thank you.

8 MR. POEHLER: Thanks. Okay. Next slide, 9

please. Okay. So this is talking more about non-10 mandatory appendix HBB-T and trying to break that down 11 a little bit and just discussing the characteristics 12 and also the evaluation methods for some of these 13 deformation-controlled quantities. So you have limits 14 for strain accumulation of 1 percent average, 2 15 percent linearized bending, or 5 percent maximum.

16 Also, creep and fatigue have to be -- creep and 17 fatigue and buckling have to be evaluated.

18 And as we mentioned before, these things 19 are typically driven by secondary stresses. The 20 right-hand side of this slide talks about the 21 different analysis methods that are available. These 22 include elastic, inelastic, and elastic perfectly-23 plastic analysis.

24 For elastic and inelastic analyses, they 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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28 can be applied to all materials. And the rules are in 1

HBB-T, elastic analysis thought to be bounding while 2

inelastic analysis is thought to be more accurate.

3 And inelastic analysis, there are no material models 4

currently in Division 5 for those inelastic analyses.

5 And then elastic perfectly-plastic 6

analysis supplies right now only to a subset of the 7

materials. And those rules are in two code cases.

8 And it's also considered a bounding analysis. Next 9

slide, please.

10 Okay. So now I'm going to talk about how 11 creep fatigue is evaluated. So creep fatigue is 12 assessed based on the interaction diagram which you 13 see on the left there. A life fraction of creep 14 damage and a usage fraction for fatigue damage are 15 determined separately.

16 The fatigue use is just computed similarly 17 to fatigue for Class 1 components in Division 1, 18 except for Division 5 has its own fatigue curves. And 19 those are in terms of strain versus cycles. The 20 coordinates of these two damaged fractions are 21 compared to the interaction diagram, and they have to 22 be inside the lines to pass.

23 Different materials have different 24 allowable creep fatigue envelopes. You can see 304 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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29 and 316 have an intersection point of 0.3 on the 1

diagram which gives it a little more liberal envelope 2

while 2.25Cr and 800H have an intersection of 0.1, 0.1 3

which is more restrictive. And then 9Cr is very 4

restrictive envelope there.

5 MEMBER BALLINGER: This is Ron again.

6 Where would 617 -- sorry for keeping to harp on 617.

7 But it's the main high temperature material and it's 8

not on here.

9 MR. POEHLER: Yeah. So this is just an 10 example. But, yeah, I can't tell you off the top of 11 my head what the interaction diagram looks like. But 12 we can --

13 DR. SHAM: 617 is 0.1, 0.1, Ron.

14 MR. POEHLER: 0.1, 0.1. Thanks, Sam.

15 CHAIR RICCARDELLA: 0.1? Okay. Thank 16 you.

17 MR. POEHLER: Okay.

18 CHAIR RICCARDELLA: So it'll be the middle 19 of the three curves.

20 MR. POEHLER: Thanks. And we'll talk more 21 about how creep damage is assessed on the subsequent 22 slides. Next slide, please.

23 CHAIR RICCARDELLA: Well, so the red and 24 blue data points on this slide are a pass versus a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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30 fail. Is that the idea?

1 MR. POEHLER: Yeah, I think those are just 2

examples.

3 CHAIR RICCARDELLA: Yeah.

4 MR. POEHLER: I think the blue one would 5

pass for stainless steel, and the orange one would 6

fail.

7 CHAIR RICCARDELLA: Yeah, go it.

8 MR. POEHLER: Let's go to the next slide.

9 Okay. So this slide goes into a little more detail 10 about how the creep damage fraction is determined in 11 the creep fatigue assessment. So creep damage for 12 different cycle types is based on stresses, and it 13 accounts for stress relaxation.

14 The upper right figure shows a schematic 15 of a stress relaxation profile. And the isochronous 16 stress-strain curves are used to determine the amount 17 of stress relaxation. The stress rupture curves are 18 used to obtain the rupture time associated with the 19 relaxed stress for the cycle type in question.

20 The lower right graph shows the stress 21 rupture curves for Alloy 617. And the time to rupture 22 represents the denomination -- denominator and the 23 creep damage term. Welds have a stress rupture factor 24 to account for the reduced rupture strength of welds 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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31 compared to the corresponding base metal. And that's 1

called out in HBB-T-1715 which requires supplying this 2

to the stress rupture curves when you did a creep 3

damage calculation.

4 CHAIR RICCARDELLA:

But

Jeff, the 5

relaxation only occurs for deformation control 6

stresses, right?

7 MR. POEHLER: Right. It's not -- you 8

don't take that into account for primary --

9 (Simultaneous speaking.)

10 CHAIR RICCARDELLA: Right. And -- okay.

11 MR. POEHLER: Okay. Let's go to the next 12 slide. Okay. Yeah, so a little bit about the 13 buckling rules, there's different buckling limits 14 depending on whether creep is significant or not and 15 also whether the buckling is either strain-controlled 16 or load-controlled. So load-controlled buckling is 17 characterized by continued application of applied load 18 in the post-buckling regime leading to failure, such 19 as, for example, collapse of a tube under external 20 pressure.

21 Strain-controlled bucking is characterized 22 by an immediate reduction of strain-induced loading 23 upon initiation of buckling and by the self-limiting 24 nature of the resulting deformations. Even though its 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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32 self-limiting, strain-controlled buckling must be 1

avoided to guard against failure by fatigue, excessive 2

strain, and interaction with load control instability.

3 So figures like the ones shown here provide time-4 temperature combinations below which the time 5

independent buckling limits may be used.

6 And this figure is an example provided for 7

one geometry. There's figures for several different 8

geometries in Division 5. For conditions where 9

strain-controlled and load-controlled buckling may 10 interact or significant elastic follow-up may occur, 11 the load factors for load-controlled buckling are also 12 to be used for strain-controlled buckling.

13 And the term, elastic follow-up, refers to 14 a situation where only a small portion of the 15 structure undergoes inelastic strains while a major 16 portion of the structure behaves in an elastic manner.

17 And in these cases, certain areas may be subjected to 18 strain concentrations due to elastic follow-up of the 19 rest of the connected structure. The next slide.

20 Okay. I'm going to talk a little bit about -- more 21 about the elastic perfectly-plastic or EPP analysis.

22 So it's a methodology for analysis of 23 deformation-controlled quantities. It's implemented 24 via two code cases as I mentioned. There's one code 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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33 case for strain limits and one code case for creep 1

fatigue.

2 The staff is reviewing Rev. 0 of the code 3

cases which only cover Type 304 and 316 stainless 4

steel. However, Grade 91 and Alloy 617 are covered by 5

revisions of those code cases. EPP is intended to be 6

easier to implement than inelastic analysis, but it 7

removes some of the over-conservatism of elastic 8

analysis methods.

9 And some of the advantages include that 10 you don't have to do stress classification. You can 11 apply it to any geometry or loading. It accounts for 12 redundant load paths, and it's simpler to implement.

13 It's based on finite element results at 14 integration points. So there's no linearization of 15 stresses. And it uses the concept of a pseudo yield 16 stress which is determined by trial and error.

17 The trial value will be the lower of the 18 yield strength or the stress to cause an -- Stress X 19 to cause inelastic strain in the time interval as 20 determined from the isochronous stress-strain curves 21 in Section III, Division 5. And that X is the -- so 22 if the component then fails, basically doesn't shake 23 down to elastic action, then you pick a different X, 24 basically. So it's kind of a trial and error process.

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34 Okay. Next slide, please. Question? No?

1 So, now I'm going to give just a little 2

more background on inelastic analysis methods. So the 3

code doesn't provide inelastic material models right 4

now. So currently this would be left to the designer 5

if he were using the 2017 edition.

6 Or actually, yeah, the code committees are 7

working on developing these models. There is some 8

historical experience from the Clinch River breeder 9

reactor with inelastic analysis of high temperature 10 reactor components. And this experience showed 11 inelastic analysis is the least ever conservative of 12 the Division 5 options.

13 It can be necessary in critical locations 14 where designed inelastic analysis is too conservative 15 to produce a reasonable design. And finally, the 16 current status of development of material models for 17 inelastic analysis in the code is that unified 18 viscoplastic constitutive models for 316H stainless 19 steel and Grade 91 have been developed. And an action 20 to add Grade 91 -- the Grade 91 model to the code has 21 just been balloted. Next slide, please.

22 Okay. So moving on to the Class B rules.

23 So the Class B rules for low temperature components 24 are essentially the same as those for Section III, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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35 Division 1, Class 2 components. For Class B high 1

temperature components, the rules do take creep into 2

account but are simplified compared to the Class A 3

high temperature rules.

4 And there's a lot more materials allowed 5

for Class B high temperature components than for Class 6

A high temperature components. Creep can be neglected 7

for components with non-negligible creep. There is a 8

Mandatory Appendix HCB-III that defines times and 9

temperatures where creep effects can be neglected.

10 Next slide.

11 A little more about the Class B rules.

12 Basically, they extend the design methodologies of 13 Division 1, Class 2 to higher temperatures. These are 14 designed by rule approach. They don't use the design 15 lifetime concept.

16 Allowable stresses are based on 17 extrapolated 100,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> creep-rupture properties 18 which is similar to Division 1. And fatigue damage 19 from cyclic service is addressed only for piping with 20 creep effects. A stress range reduction factor is 21 used, similar to Division 1, Class 2, but the factors 22 are reduced to account for elevated temperatures.

23 Next slide.

24 So for metallic core supports, you have 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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36 low temperature rules in HGA which are essentially the 1

same as those in Division 1 for core supports. And 2

then for elevated temperature metallic core supports, 3

the rules are essentially the same as those for Class 4

B. I couldn't say Class -- I mean, I meant Class A, 5

Class A elevated temperature components, including the 6

same allowable materials and stresses. Next slide.

7 Okay. Now moving on to construction rules 8

for nonmetallic components. So Division 5 is unique 9

in that it provides rules for nonmetallic components, 10 including both graphite and composites. Graphite 11 materials are used mainly in core components in 12 certain advanced reactor designs due to their 13 excellent neutron moderation properties.

14 Rules for composites were added in 15 Division 5 for the 2019 edition. In the 2017 edition, 16 the rules for composites were listed as in the course 17 preparation. So the staff did not review those, the 18 rules for composites. Next.

19 CHAIR RICCARDELLA: What do you mean by 20 composites? Graphite is one?

21 MR. POEHLER: No, I think --

22 DR. SHAM: They are the C/SiC composite or 23 24 DR. WINDES: C/SiC and carbon-carbon.

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37 MR. POEHLER: Silicon carbide maybe.

1 DR. WINDES: Yes.

2 MR. POEHLER: Yeah.

3 DR. WINDES: Yeah, silicon carbide matrix, 4

silicon carbide fiber as well as carbon fiber and 5

carbon matrix. So carbon-carbon and C/SiC.

6 CHAIR RICCARDELLA: Okay. Thank you.

7 MR. POEHLER: Thank you. Next slide, 8

please. So now I'm going to talk about some of the 9

characteristics unique to graphite that provides a 10 little background to help understand the provisions of 11 Division 5 for graphite design and materials. So some 12 of these include the fact that there's no single 13 nuclear grade of graphite. Therefore, we can't design 14 around a specific nuclear grade as we can for metals 15

-- metallic materials.

16 Graphite is heterogeneous by nature and 17 contains significant pores and cracks. Graphite is 18 not ductile. It has brittle or quasi-brittle fracture 19 behavior. And so the graph here on the right of this 20 slide shows an example of turnaround which is 21 basically you have a volume change initially with 22 increasing neutron dose where the volume shrinks up to 23 a certain dose and it begins to expand. And the 24 material's behavior is completely different before and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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38 after the turnaround does is accumulated. Next slide, 1

please.

2 MEMBER BROWN: This is Charlie Brown 3

again. Could you go back to that graphite slide?

4 MR. POEHLER: Yeah, let's go back.

5 MEMBER BROWN: I'm not a materials guy, 6

just trying to make sure I'm educated with the 7

advanced reactor somewhat. Very graphically, describe 8

the negative aspects of graphite in the application of 9

the advanced reactors. Is that going to result or do 10 you think it would result in a change of their seismic 11 response? Do we have to change seismic rules to allow 12 these things -- these materials to be used?

13 MR. POEHLER: That's a good question. I 14 would probably maybe ask Will Windes if he could talk 15 to that a little bit.

16 DR. WINDES: Yeah, I think it -- first of 17 all, I think it depends upon the design. So as you 18 can see, you're looking at maybe a 5, 6, 7 percent 19 volumetric change macroscopically at the most for 20 whatever grade of graphite. Sometimes you're only 21 looking at something like one -- a half to one percent 22 volumetric change.

23 So, dependent upon the grade of graphite 24 that you use, the design that you have, then, yeah, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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39 you are going to have to maybe consider something like 1

a seismic. But again, it's going to be very, very 2

specifically design oriented. Does that --

3 (Simultaneous speaking.)

4 MEMBER BROWN: Go ahead.

5 MEMBER KIRCHNER: Charlie, this is Walt.

6 These kind of -- this curve we're looking at here 7

certainly was a big factor in the Fort St. Vrain 8

design which used prismatic graphite blocks. And so 9

yes, seismic is one of the issues. Bypass is another 10 issue that was a concern. And subsequent designs of 11 the modular HTGRs that were using prismatic blocks 12 instead of pebbles made various design-specific 13 changes.

14 For example, they put, like, a cap. And 15 I'm not describing it very well. But instead of just 16 having graphite blocks -- prismatic blocks stacked on 17 each other, they had a little crown that went over --

18 in the advanced designs over the graphite, a block 19 that was below it so that they didn't have wobbling, 20 so to speak, under flow and then having bypass and 21 other kind of issues also and structural stability to 22 deal with things like seismic loadings and such.

23 CHAIR RICCARDELLA:

So that's the 24 shrinkage concern. So that's in the beginning of this 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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40 radiation effect when the volume change is actually 1

shrinkage?

2 MEMBER KIRCHNER: Yeah, it's shrinkage, 3

the first feat that they had to deal with. I don't 4

know that they were looking at exposures that got back 5

up above the curve where it changed. I think in the 6

end reactor, they had those kind of problems, though.

7 That was a production reactor for the weapons program.

8 But they had, I think --

9 (Simultaneous speaking.)

10 MEMBER KIRCHNER: -- entire fluences in 11 that. And they did cross the curve that you're 12 looking at.

13 MEMBER PETTI: So in general, though, for 14 some of these reactors, the design criteria is that 15 you don't design beyond the minimum shrinkage. Others 16 will talk about designing up to the point that you go 17 back to zero. Nobody talks about designing in the 18 swelling region above zero.

19 The other thing is that the grade 20 sometimes can be used. This is in the middle of the 21 core where the fluence is the highest. The support 22 structure, the fluences are much, much lower.

23 Sometimes other grades are used. It's not all the 24 same grade in the core. So there's a lot of design 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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41 considerations here. And the HTGR experts are well 1

aware of these things.

2 MEMBER REMPE: So Dave, back in the days 3

of General Atomics, they -- help me remember. Wasn't 4

it called H-451 or something --

5 MEMBER PETTI: Yes.

6 MEMBER REMPE: -- is what we had.

7 MEMBER PETTI: Yes.

8 MEMBER REMPE: And had they -- and I know 9

that source is no longer available. Have all these 10 designers -- because there's quite a few folks 11 thinking they're going to do something with a graphite 12 reactor, for the fuel or for the moderator or 13 whatever. And have they identified sources? Where 14 are they?

15 MEMBER PETTI: Yes, so you see all the 16 data there. All major grades that are available with 17 all the major vendors have been tested in the DOE 18 program. And these are -- let's call them new grades.

19 They all -- you could tie them back to the 20 old grades like H-451. There was an equivalent German 21 graphite grade. And so there's a lineage, if you 22 will.

23 But there's a lot of grades out there 24 besides the old German and the old H-451 which was the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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42 American. There's Japanese grades now. China is 1

trying develop their own grades.

2 MEMBER REMPE: How do they compare if I 3

look at this --

4 (Simultaneous speaking.)

5 MEMBER PETTI: At least as good as the 6

historic.

7 MEMBER BALLINGER: And I think IG 110 is 8

probably better.

9 MEMBER PETTI: Well, IG 110, yeah, that's 10 the Japanese grade. But if you look at the American 11 grade that replaced H-451, it's at least as good, if 12 not better.

13 (Simultaneous speaking.)

14 DR. WINDES: I'm sorry. Yeah, I was going 15 to say, so Joy, just to give you an idea. The PCEA, 16 the blue square, was Graphtec International's attempt 17 to duplicate after 40 years the old H-451 recipe. And 18 we actually had legacy H-451 graphite that we put into 19 their first two capsules of the AGC experiment and did 20 a direct one-for-one comparison between H-451 and 21 PCEA. And at least from an irradiation response and 22 behavior standpoint, they lay on top of each other so 23 well that you can barely distinguish between H-451 and 24 PCEA. And that's --

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43 MEMBER REMPE: And is there a huge amount, 1

Will? I'm sorry to interrupt. But is the amount 2

large? Do they have a huge source? They're not going 3

to have to do this because they're going to run out 4

again or something?

5 DR. WINDES: No. So that's -- and that 6

his one of the questions that's going on right now.

7 And I'm sure that the NRC is going to be involved in 8

that is that the whole issue of source, let's face it.

9 You're not going to be able to duplicate graphite-like 10 metal because you don't take it down to the atomistic 11 composition.

12 You take it down to basically its 13 molecular airmatic (phonetic) ring structure. And 14 that is dependent upon where you get your source 15 material. So even if you dig the same coal out of the 16 same coal mine or pump it out of the same oil well, 17 the farther down you go in that coal mine or in that 18 well, you're going to have a geologic change to the 19 source material.

20 But with that said, I mean, everybody 21 knows this. It's out in the open. This is a 22 potential issue and weakness. But with that said, the 23 graphite suppliers are well versed and have a lot of 24 experience in determining and correcting and changing 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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44 the formulas so that you get the same response because 1

this has been going on since basically we made 2

synthetic graphite for over 100 years. People want a 3

consistent material. And so they have experience to 4

do that.

5 And that's what the experiment with PCEA 6

was. It was a completely different source of raw 7

material, completely different facility, absolutely no 8

people that had done H-451 and made it. And yet after 9

40 years of laying dormant, they were able to 10 resurrect the recipe and show that they could produce 11 a material that had the same characteristics as a 12 material that had been produced 40 years previously, 13 without the same material, without the same source 14 material, or coke source or anything else.

15 So that's a question that's being debated.

16 I think that most people believe that we can go in and 17 create a grade of graphite that is consistent 18 throughout time. So if you wanted to have a second, 19 third, or fourth core replace the components, I can 20 tell you that the graphite community is very confident 21 that the suppliers can produce a grade even 20 or 30 22 years later that is consistent with that first core.

23 Does that make sense?

24 MEMBER REMPE: Thank you. Yeah, thanks, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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45 Will. It's good to talk to you again, even if it's 1

virtually.

2 DR. WINDES: Yeah, that too.

3 CHAIR RICCARDELLA: Walt, this is Pete.

4 MEMBER KIRCHNER: This is Walt Kirchner.

5 How well have they done with a Great Lakes carbon 6

supply? One of the big issues is neutronics and 7

impurities. So how well are they doing when they 8

replicated the H-451? How well did they do on 9

neutronic impurities?

10 DR. WINDES: Oh, that's pretty --

11 MEMBER KIRCHNER: It's a side question, 12 but it's an important one.

13 DR. WINDES: Yeah. No, that's -- the 14 purification process is actually probably a lot 15 better. One of the things that -- while the nuclear 16 industry sort of stayed still and dormant in this area 17 and we really haven't pushed the technology, the IT 18 industry has. And in fact, just as a little anecdote 19 to answer this question indirectly, when we went in 20 and did a quality assurance inspection on one of the 21 graphite suppliers and we told them that this was 22 going to be a nuclear quality assurance inspection and 23 they were all revved up, they called us back 24 afterwards and they said, man, that was easy.

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46 If you want to see what specs are and 1

getting the impurity levels down, you got to go 2

through an IT inspection. So what's happened in the 3

last 20 to 25 years is that the IT, specifically the 4

silicon chip and all of the computer and solar panel 5

folks, they have come in and they have progressed the 6

purification process to the point that we never could 7

have in the past with the nuclear program. So yeah, 8

it's much better even then than we had in the past.

9 MEMBER PETTI: Walt?

10 MEMBER KIRCHNER: Thank you. That's good 11 to know. Okay.

12 MEMBER PETTI: Just so you know, these 13 samples that are irradiated, they can be contact 14 handled.

15 DR. WINDES: Oh, yeah.

16 MEMBER PETTI: They're not very hot at 17 all.

18 DR. WINDES: Yeah.

19 MEMBER PETTI: That may not have been the 20 case years and years ago.

21 MEMBER KIRCHNER: Well, back in the '80s, 22 when Great Lakes Carbon was no longer a source of 23 supply, what I was doing was mining older logs.

24 MEMBER PETTI: Yeah.

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47 MEMBER KIRCHNER: But that didn't bode 1

well for the MHTGR program in circa the '80s. So this 2

is encouraging news.

3 DR. WINDES: Oh, yeah, yeah. And like I 4

said, it's a lot more -- the purification process is 5

a lot more sophisticated than it ever was for the old 6

327 and the H-451 graphite grades.

7 CHAIR RICCARDELLA: Walt Kirchner, in your 8

initial comments, you distinguished between prismatic 9

core elements versus pebble bed. Could you give a 10 little more on why it is that distinction? Is it less 11 critical in a pebble bed reactor?

12 MEMBER KIRCHNER: Yeah, it's much less 13 critical. Dave could speak to it better than I could.

14 But you don't have such a large structure as you --

15 those prismatic blocks were typically about a meter 16 high, 12 or 14 inches across the flats in a hex 17 configuration. So you've got an actual structure that 18 is in both a thermal and a radiation field that varies 19 both -- in all dimensions. So that creates a lot more 20 challenges for the core designer than dealing with a 21 nice hard pebble.

22 MEMBER PETTI: But I will say, though, 23 that the reflector of a pebble bed is quite a 24 challenge structurally. There are different issues.

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48 It's keyed together.

1 Think of the -- the prismatic is kind of 2

like Lego blocks with straps around them. So it's 3

thermomechanically easier than if you look at the 4

design of the reflector in a pebble bed. And 5

particularly, even the support, you've got to have --

6 you've got to hold core up and you've got to let the 7

pebbles through. It's quite the challenge.

8 DR. WINDES: And I will point out that the 9

image in the lower left-hand corner, that is some of 10 the outer reflector bricks that were designed by the 11

-- for the pebble bed modular reactor, the PBMR in 12 South Africa. And if you look at that, you can see 13 what Dave's talking about. They're keyed together, 14 and they have to be interlocked just basically to 15 support those pebbles that are inside there.

16 And then from a seismic standpoint -- and 17 this is why composites is being considered. But from 18 a seismic standpoint, they had silicon carbide or 19 carbon-carbon belts that wrapped around the core 20 purely for seismic considerations. And they basically 21 provided a tensile restraint during seismic events --

22 potential seismic events.

23 CHAIR RICCARDELLA: Understand. Okay.

24 Thank you.

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49 MEMBER BROWN: Can I ask my question in a 1

different way? You all buried me in prismatics and 2

everything else. Let me put this more practically.

3 All the stuff here, it's very brittle. I know people 4

have made advances like you all commented on. But I'm 5

thinking about it in a long-term application, for 6

example, conventional reactor fuels as we has today.

7 We had an earthquake at North Anna. And 8

within a short period of time after that, it rode 9

through. It started back up and had no -- and 10 operated as if nothing ever happened. If you have a 11 seismic event of that nature with a graphite-type 12 moderator, is there a concern that you'll be able to 13 go right back to operation? Or are you going to have 14 to go in and do something in the plant?

15 MEMBER PETTI: There will be a safe 16 shutdown earthquake, and they will have to design it, 17 right?

18 MEMBER BROWN: I'm not worry about safe 19 shutdown, Dave. I'm talking about below safe 20 shutdown.

21 CHAIR RICCARDELLA: And that would be OBE, 22 an operating basis earthquake.

23 MEMBER BROWN: Yeah, and North Anna rode 24 through that and nobody blinked. They kept on 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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50 trucking. Will graphite be able to do that?

1 MEMBER PETTI: It will have to be.

2 MEMBER BROWN: How do we prove that?

3 MEMBER PETTI: Through the analysis.

4 CHAIR RICCARDELLA: That's what this code 5

is all about.

6 MEMBER BROWN: Okay. All right. You've 7

answered -- we're not there yet is what you're 8

fundamentally telling. There's a lot of work to be 9

done to prove that we'll ride through that similarly.

10 I'm just looking at long-term performance. That's all 11 I'm --

12 MEMBER PETTI: Yeah, I mean, Fort St.

13 Vrain had an earthquake they had to survive, as we 14 know, so --

15 MEMBER BROWN: Yeah, how long was it in 16 operation? Or how long it was built before they shut 17 it down? That's a big difference.

18 MEMBER BROWN: How long did it --

19 (Simultaneous speaking.)

20 MEMBER BALLINGER: When it was above water 21 or under water?

22 MEMBER BROWN: How long did it operate?

23 MEMBER KIRCHNER: Well, Charlie, this is 24 Walt. The right answer here is putting Fort St. Vrain 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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51 aside which was we can go through all the reasons it 1

was shut down. But when they -- certainly when they 2

designed it, they designed for a 40-year life.

3 And if I remember correctly, they had 4

shake table tests for their core and reflector 5

structure so that they could convince themselves that 6

from the stress and from the mechanical design 7

standpoint that they configuration would meet both the 8

OBE and SSE requirements. So I'm confident that they 9

can design a 40-year core, and if it were exposed to 10 an event like in Virginia, if it's below the safe 11 shutdown or OBE limits, I'm confident that they would 12 be able to restart the reactor.

13 MEMBER BROWN: Okay. That's all. That's 14 what I'm --

15 (Simultaneous speaking.)

16 MEMBER BROWN: You're an expert. You all 17 18 DR. WINDES: May I say one thing?

19 MEMBER BROWN: Pardon?

20 DR. WINDES: May I say one thing, please?

21 MEMBER BROWN: Sure, yes, yes.

22 DR. WINDES: Yeah. Just so -- let me ask 23

-- let me answer it in two different ways, sort of a 24 Part 1, Part 2. First and foremost, from a material 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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52 science standpoint, graphite, these components are 1

fairly massive. Putting a crack through those or even 2

chipping an edge off of them so that they may --

3 during a seismic event so that they would not -- the 4

Legos would not fit into each other would be -- from 5

a material science standpoint, would be highly, highly 6

unlikely.

7 There would have to have been a major flaw 8

near the edge that was undetected. So if the rules --

9 design rules are followed and all of the inspections 10 are followed, there should be no reason for these 11 things to -- the individual components to stay 12 completely and totally stable. The graphite is robust 13 enough to do that. We're not making this out of 14 glass. Graphite is a lot more forgiving. So from a 15 material standpoint, that's not a problem.

16 The second part is, is that remember that 17 the core is made up of individual stacked components.

18 So they're not rigid. So if there is a crack that 19 forms in one of these components, the real question 20 is, who cares, because we already have cracks.

21 We're stacking individual elements 22 together and the gaps between them is significant.

23 They are huge cracks if you want to think of them that 24 way. So if a small crack occurs, that's not really a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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53 problem.

1 Even if a small chip occurs, that's not 2

really a problem. What's really a problem -- and this 3

is why I answered your initial question with, well, it 4

depends upon the design. The real issue is, can you 5

shut down the reactor in a safe and timely manner?

6 Can you keep the safe -- the fuel safe? Can you keep 7

the safe operation of the core?

8 And the answer to that is across the 9

Atlantic and that is with the AGR reactors. They have 10 done extensive testing of their core's shake tables, 11 a quarter size, full size reactor cores with a quarter 12 size on gigantic shaped tables. And they have gone in 13 and done up to the maximum expected seismic events 14 that they have in England and found absolutely no 15 problems with their design.

16 And the last thing I will say is that 17 every single brick right now in the UK is cracked, has 18 at least one, if not two through cracks. And yet they 19 can still operate their reactors safely. And they 20 have done so for 20-plus years.

21 MEMBER REMPE: But Will --

22 DR. WINDES: So again -- what?

23 MEMBER REMPE: -- aren't those cracks why 24 they're shutting down the UK reactors prematurely?

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54 DR. WINDES: No, not really. What they're 1

doing is they're extending the life beyond what the 2

original design life was, which was about 20, 25 3

years. They're actually extending it beyond.

4 And as a consequence -- and the real 5

problem is, is that these -- and it all gets back to 6

the design of the actual cores and what you intended.

7 These -- their cores are so keyed and interlocked that 8

they're literally sort of a one shot and they're done.

9 And there's no way you could go in there and pull out 10 a cracked element -- or excuse me, reflector element 11 and pull it out and replace it.

12 You have to completely disassemble the 13 entire core. And so as a consequence from the 14 economic standpoint, you can't do that because it's so 15 keyed together. Cracks don't really matter to them.

16 They've operated safely for decades with cracked 17 components. But they don't really care because the 18 core is designed to actually withstand that kind of 19 phenomenon. So again --

20 MEMBER REMPE: To say they don't really 21 care, I know that there's been discussions for decades 22 about those cracks.

23 DR. WINDES: They care immensely about 24 that. But does it -- is it a critical safety problem?

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55 And the answer is no. They care immensely about it.

1 And there's been hundreds of pounds -- millions of 2

pounds that have been proposed -- or excuse me, been 3

worked on this issue.

4 So they care immensely about this. But 5

what the question that comes down to is, can they 6

operate their reactor cores safely with cracked 7

components? And the answer is yes, not only in their 8

models, not only in their analysis, but through pure 9

experience. In the last 20 years, they've had a 10 number of issues and it's never compromised the safe 11 operating envelope of a single one of their reactors, 12 even though everybody knows they are cracked bricks.

13 So cracked bricks is not necessarily a 14 stopping of the entire reactor consideration. So you 15 have to have that knowledge as well when you're 16 designing these cores. And I apologize. I've taken 17 up a lot of time in this.

18 MEMBER BROWN: No, don't apologize. I'm 19 not a -- obviously not an expert on graphite, and I 20 know we're going to have a lot discussions later. But 21 this has been an excellent discussion. I appreciate 22 your time and the patience --

23 DR. WINDES: Oh, no. Thank you for 24 listening. I'll talk all day about this.

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56 CHAIR RICCARDELLA: A question on seismic:

1 the UK is not really a high seismic region, is it?

2 DR. WINDES: No, it's not. What they're 3

talking -- I think, if I'm not mistaken, they're 4

taking about something on the order of five, five and 5

a half is what they're really, truly expecting. But 6

I believe the original question was something that was 7

not a catastrophic shutdown event but basically a 8

small event that could have a restart. The UK is a 9

perfect example for something like that.

10 CHAIR RICCARDELLA: An OBE, but there's 11 something that gives metallic structures almost an 12 inherent -- makes them inherently forgiving to seismic 13 loads because most seismic design work is now with 14 linear analysis and you're worried about resonance at 15 certain frequencies. And as soon as you exceed the 16 yield strength in a metallic component, you get a 17 little bit of yielding that introduces stamping that 18 changes where you are on the resonance curve. And so 19 the loads go down compared to what the elastic 20 analysis would predict.

21 I'm not sure if that same phenomenon works 22 in graphite -- in a graphite -- on the slide, it has 23 graphite. It's not ductile. It's brittle or quasi-24 brittle. To me, it's almost like masonry structures 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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57 don't respond real well to earthquakes compared to 1

steel or wood frame structures.

2 DR. WINDES: True, but the design is 3

completely different. In a metallic, you would have 4

a pressure retaining structure whereas in the graphic, 5

we are not a pressure retaining vessel.

6 CHAIR RICCARDELLA: Yeah, yeah.

7 DR. WINDES: And so as a consequence, the 8

entire design requirements and the function of the 9

graphite is not to go in and withstand a cracking 10 event. It's to maintain the structural integrity of 11 the core.

12 CHAIR RICCARDELLA: Okay.

13 MEMBER BALLINGER: Yeah, this is Ron. I 14 mean,Section V does not account for -- there's an 15 explicit thing in there. It says, we don't count for 16 corrosion. The equivalent for graphite if there's no 17 water in the system is probably wear. Am I correct?

18 Erosion?

19 DR. WINDES: Yeah. Well, wear and erosion 20 is something that we're considering. But quite 21 frankly, it depends on the molten salt or if you have 22 a gas cooled environment.

23 MEMBER BALLINGER: Yeah, yeah.

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58 and erosion is going to be something. For a gas 1

cooled reactor, the only real issue is probably dust 2

entrained high velocity gas in some of these regions.

3 But again, the wear and erosion is probably not 4

something that's really where we're really worried 5

about.

6 I think really what the main issue is the 7

-- and this is why it's in the design rules itself is 8

the irradiation effects of the graph on the graphite.

9 So unlike metals where it basically sort of bottoms 10 out, the graphite has this sort of dynamic response 11 and behavior. And it changes, as you can see, as a 12 function of dose.

13 And that's why turnaround is so critical 14 and important. Once you figure out where your 15 turnaround is, then you can predict and understand 16 what the behavior is going to be like. But it's a 17 dynamic response to the irradiation and the radiation 18 temperature. That's why it's in the design rules and 19 not in Section VIII.

20 MEMBER BALLINGER: Yeah, I remember 21 sitting in Arkal Shenoy's office where he had a 22 graphite block that was tested for the Fort St. Vrain 23 reactor. And that graphite block after exposure to a 24 test loop had about an inch of wear off of one of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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59 those blocks.

1 DR. WINDES: Right, yeah. It's both soft 2

and hard.

3 MEMBER BALLINGER: Yeah.

4 DR. WINDES: My machinist in the back 5

machining samples both loves and hates graphite at the 6

same time. So yeah, it's beautiful. It's easy. It 7

cuts and then it dulls as cutting tools like nothing 8

else, so yeah.

9 MEMBER KIRCHNER: Yeah.

10 DR. WINDES: Very, very weird.

11 MEMBER KIRCHNER: Ron, this is Walt. The 12 design challenges are quite a bit different than using 13 a metal core. And the picture in the lower left is 14 illustrative of some of the things you would worry 15 about. You don't want excessive wear creating dust 16 and contamination in the primary circuit. You don't 17 want large bypass because of the volumetric shrinkage 18 there before you get to turnaround.

19 You have to worry -- probably the biggest 20 seismic worry is not the blocks as Will was saying, 21 cracking and such. The biggest worry is alignment so 22 that you can ensure that if you're using control rods, 23 you can get the controls rods inserted and achieve a 24 safe shutdown condition. So it's a different set of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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60

-- there are similar issues but a different set of 1

problems that you deal with, especially for the gas 2

cooled --

3 DR. WINDES: Yeah, that's where the --

4 that's why I was mentioning where because that changes 5

-- if you have a lot of wear, it might change the 6

seismic response.

7 CHAIR RICCARDELLA: Yeah, so this is a 8

very interesting discussion, but I think we have to 9

move on. We've got about five more slides, I think, 10 in the overview part of the Section III, Division 5.

11 And then I'd like to take a break. And then we'll get 12 into the staff -- the comments on the staff 13 endorsement of Section -- of Division 5.

14 MR. POEHLER: Thanks, Pete. Okay. Yeah, 15 so this -- now moving on, talking about some of the 16 code considerations here with graphite. Because all 17 graphite is brittle and contains flaws as we 18 discussed, core components need to be designed to 19 accept some amount of cracking. The upper right 20 figure shows some internal flaws in graphite.

21 So because of these characteristics, a 22 probabilistic versus deterministic design approach 23 needs to be used because deterministic is generally 24 too limiting for brittle material like graphite. So 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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61 distribution and possible strengths in the material is 1

needed for a material like this. And a probability of 2

failure in components is based on inherent strength of 3

graphite grades and applied stresses during operation.

4 So the figure on the left kind of shows 5

distributions of loading on the left-hand curve on 6

that figure and the distribution of material strength 7

on the right-hand curve on that figure. The overlap 8

of those two curves represents the reliability of the 9

part. So let's move on. Next slide.

10 MEMBER BALLINGER: Is there a Weibull 11 modulus spec on this stuff?

12 DR. WINDES: Yes.

13 MR. POEHLER: I don't know the answer to 14 that. I would --

15 DR. WINDES: Yes, I think there -- that's 16 what you're seeing right here is viable strength 17 curves. And that's --

18 MEMBER BALLINGER: Okay, okay. That's 19 what I thought.

20 DR. WINDES: Yeah.

21 MR. POEHLER: Okay. So those slides are 22 talking about the structural integrity assessment 23 methods that are in Division 5 for graphite 24 components. The upper -- or the figure on the right 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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62 here shows typical material testing curves used to 1

derive failure probabilities, tensile strength versus 2

failure probability. So the methods that are in the 3

code for assessment, they're three basic methods, the 4

simplified assessment which is a

simplified 5

conservative method based on ultimate strength derived 6

from Weibull statistics.

7 The full assessment is a more detailed 8

assessment that takes into account

stresses, 9

temperatures, a radiation history, chronic -- and 10 chronic oxidation effects. Weibull statistics are 11 used to predict failure probability. The maximum 12 allowable probability of failure is determined for 13 three structural reliability classes which related to 14 safety function.

15 And so those three classes are shown in 16 the table here along with a maximum probability of 17 failure allowed. And then finally, design by test is 18 also allowed by the code. And that involves full 19 scale testing to demonstrate that failure 20 probabilities meet the criteria of a full analysis.

21 I'd like to point out the graphite rules are a 22 process. The designer can't just pick a pre-approved 23 material. The designer has to demonstrate their 24 specific graphite grade selected will consistently 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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63 meet the component requirements.

1 CHAIR RICCARDELLA: So this SRC-1, SRC-2, 2

are they somehow analogous to Class A, Class B 3

components?

4 MR. POEHLER: I believe so. I think it 5

relates. You would designate that based on the safety 6

significance of --

7 CHAIR RICCARDELLA: Yeah, okay.

8 MR. POEHLER: -- the consequences for the 9

year of the --

10 CHAIR RICCARDELLA: Right. And failure 11 doesn't necessarily mean failure of the structure. It 12 just means cracking?

13 MR. POEHLER: Correct, the probability of 14 a through crack.

15 CHAIR RICCARDELLA: Okay.

16 MR. POEHLER: Okay. Next slide, please.

17 So yeah, so anyway, this is addressing some of the 18 special considerations in the design of graphic 19 components, and those include oxidation, irradiation 20 and abrasion, erosion which Division 5 says should be 21 addressed. This figure kind of shows how these 22 special considerations can shift both the loading 23 distribution and the strength distribution in either 24 direction which would change the overlap area for the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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64 two distributions. So degradation can change -- or 1

radiation can increase strength or typically increase 2

the strength.

3 High temperature can increase strength.

4 Oxidation would decrease strength. Molten salt may 5

decrease strength. Irradiation changes also changes 6

the stress loading on the part.

7 Dimensional changes can increase stress.

8 But irradiation creep on the other hand can relieve 9

stress. So the stress distribution curve here on the 10 left could shift either way due to irradiation. And 11 those shifts could change this overlap here. So those 12 have to be considered. Okay. Next slide, please.

13 CHAIR RICCARDELLA: On the previous slide 14 where you were talking about the allowable probability 15 of failure, that really refers to the green curve, 16 right?

17 DR. WINDES: Correct.

18 MR. POEHLER: Thanks, Will. Next slide.

19 So this slide is showing the data sheet for graphite 20 which is called out in Article HHA-2-2000 material 21 data sheet forms. And this data sheet captures most 22 of the graphite degradation issues. It includes some 23 material properties or physical properties.

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65 dependence, and oxidation effects. Molten salt issues 1

aren't addressed yet in the Division 5 code. So the 2

cognizant code test group is currently working on 3

modifications to add that.

4 And let's go to the next slide. So this 5

is the summary for the overview. So just to 6

summarize, Division 5 was issued as part of the 2011 7

Addenda to the code. The design rules trace all the 8

way back to the 1960s for development of high 9

temperature rules for metallics.

10 Division 5 covers the rules for design, 11 fabrication, inspection, and testing of components in 12 high temperature reactors. And these construction 13 rules cover both metallic and nonmetallic components 14 with the rules for nonmetallic components being unique 15 among all design codes worldwide. And finally, the 16 ASME code committees are actively pursuing code rules 17 improvement and developing new technologies to support 18 Advanced Nuclear. With that, I'm going to turn it 19 over to Jordan.

20 CHAIR RICCARDELLA: Okay. So well, thank 21 you, Jeff. That was an excellent summary, and we 22 really appreciate the effort you put into it. I'm 23 going to propose now that we take a 15 minute break.

24 So we'll go into recess into, what is it, 11:10 East 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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66 Coast time. Okay.

1 (Whereupon, the above-entitled matter went 2

off the record at 10:55 a.m. and resumed at 11:09 3

a.m.)

4 CHAIR RICCARDELLA: Okay. We are 5

approaching -- it is now 11:10, and so we'll -- the 6

meeting will come to order again.

7 And I believe we've had a review of just 8

what's in Section III, Division 5, and now we'll have 9

a discussion of the NRC review and potential 10 endorsement of it. And I guess, Jordan Hoellman, are 11 you going to lead this discussion?

12 MR. HOELLMAN: That's right, Pete. I will 13 14 CHAIR RICCARDELLA: Okay.

15 MR. HOELLMAN: I will start as long as --

16 CHAIR RICCARDELLA: Thank you.

17 MR. HOELLMAN: -- everyone is ready. You 18 guys can all hear me okay, right?

19 CHAIR RICCARDELLA: Sounds good.

20 MR. HOELLMAN: All right. Awesome. So 21 good morning. My name is Jordan Hoellman. I am the 22 project manager for the endorsement effort of ASME 23 Section III, Division 5. I work in the Advanced 24 Reactor Policy Branch in NRR, and I'm excited to be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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67 here to present the staff's endorsement efforts and 1

review philosophy related to the potential endorsement 2

of Division 5.

3 So, as you know, the NRC staff is taking 4

steps to develop its regulatory infrastructure for 5

advanced non-lightwater reactors to ensure we are 6

prepared to support the review of future design 7

certifications and other licensing applications.

8 I want to take just a brief minute to 9

provide some historical context for this effort. In 10 2016, we issued the NRC Vision and Strategy for 11 ensuring or achieving non-lightwater reactor mission 12 readiness in response to the increasing interest in 13 advanced reactor designs.

14 To achieve the goals and objectives in the 15 Vision and Strategy document, the NRC staff developed 16 near-term and long-term implementation action plans or 17 IAPs. Under IAP 4, the staff intends to enhance the 18 NRC's technical readiness for potential advanced 19 non-lightwater reactor designs by applying its 20 established process for adapting its regulatory 21 framework to ensure that it facilitates the use of 22 codes and standards.

23 In 2018, ASME requested that the NRC 24 review and endorse the 2017 edition of ASME Section 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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68 III, Division 5, and the staff responded in August of 1

2018 that we were initiating efforts to endorse with 2

any limitations and exceptions, if necessary, the 2017 3

edition of the code and a new regulatory guide as one 4

way of meeting the NRC's regulatory requirements.

5 So we can move on to slide 33.

6 So the existence of robust and 7

comprehensive rules for design of high-temperature 8

reactor systems and components in the ASME code 9

endorsed by the NRC for use by prospective 10 non-lightwater reactor vendors would improve the 11 efficiency and effectiveness of the NRC's review 12 process.

13 An integral part of the framework will be 14 the endorsement of codes and standards that are 15 applicable to the construction, inspection, and 16 operation of these designs.

17 In this portion of today's briefing, we 18 will provide an overview of the review process the NRC 19 initiated for the potential endorsement of the 2017 20 edition of Division 5 and discuss some examples of 21 likely exceptions and limitations to the NRC's 22 endorsement.

23 So let's move to slide 34, please.

24 So the results -- the results of the NRC 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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69 staff's review will be compiled into two documents 1

that we are currently working to finalize and realize 2

for public comment. NUREG-2245 will document and 3

provide the technical basis for the endorsement of the 4

2017 edition of the code, as well as code cases N-861 5

and N862, which Jeff described earlier.

6 NUREG-2245 provides the technical basis 7

for the staff positions in Draft Guide 1380, which is 8

a proposed revision to Reg Guide 1.87, which is titled 9

Guidance for Construction of Class I Components in 10 Elevated Temperature Reactors.

11 The staff is currently not planning to 12 incorporate this by reference into 10 CFR 50.55(a), as 13 Section III, Division 1, is. One reason we decided to 14 do this is that the staff expects that there will be 15 continued significant revisions to Division 5 between 16 editions. And in NRC future reviews of those 17 editions, we may take a different approach to 18 endorsement.

19 By endorsing via a reg guide, our 20 endorsement, with any limitations and exceptions as 21 discussed in the reg guide, would serve as guidance 22 for a method acceptable to the staff for the use of 23 Division 5. Because we are not doing this via 24 rulemaking, an applicant can propose to use Division 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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70 5 with different limitations or exceptions, and those 1

reviews will occur in an application-specific basis.

2 The Draft Guide 1380 does include an 3

appendix, which establishes acceptable quality group 4

assignments of mechanical systems and components for 5

non-lightwater reactors acceptable to the staff for 6

the safety classification methods, including the 7

traditional means outlined in 10 CFR. 8 Using the definition of "safety-related 9

structures, systems, and components" in -- defined in 10 10 CFR 50.2, it addresses the risk-informed approach 11 outlined in 10 CFR 50.69, and it addresses the method 12 in the Nuclear Energy Institute Document 1804, which 13 is the licensing modernization project methodology, 14 which the NRC endorsed last year in Reg Guide 1.233.

15 The guidance in Appendix A is intended to 16 provide guidance on selecting an appropriate design 17 standard once the classification methods are used to 18 determine the classification of each system and 19 component. And I believe there is an ACRS briefing 20 tomorrow that will provide greater detail on the 21 licensing modernization project methodology.

22 So let's move on to slide 35.

23 So this slide just communicates the scope 24 of the staff's review of Division 5. As I previously 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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71 mentioned, Division -- well, yeah. Division 5 and 1

code cases and 861 and 862 were included in the 2

staff's review.

The staff did not review 3

non-Mandatory Appendix HBB-Y titled Guidelines for 4

Design Data Needs for New Materials. And there were 5

few portions of the 2017 edition that were in 6

preparation at the time the staff initiated our 7

endorsement effort, and we're not endorsing those 8

portions of the code at this time.

9 The staff initiated a separate effort, as 10 Jeff was describing, to endorse the Alloy 617 code 11 cases that were incorporated -- or that were approved 12 by ASME last year in 2020. The issuance of those code 13 cases represents a significant amount of work over 14 several years by the Section III subgroup on 15 high-temperature reactors.

16 The staff is reviewing these code cases 17 separately from the Division 5 endorsement effort 18 included in today's briefing, and we are considering 19 approaches to fold Alloy 617 code cases before we 20 issue the final reg guide endorsing this.

21 So slide 36.

22 As Louise was mentioning in her opening 23 remarks, the staff recognized that there was limited 24 expertise outside the ASME code developers on Division 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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72

5. To ensure an independent review, we contracted 1

with national laboratories and commercial contractors 2

for peer review on the technical adequacy of Division 3

5.

4 We held periodic teleconferences and 5

shared a collaborative SharePoint site to ensure 6

adequate resolution of technical issues raised by the 7

contractors during their independent review.

8 In addition, we contracted with Argonne 9

National Lab and Idaho National Lab because we 10 recognized that they had the foremost expertise on 11 this -- on the metallic and graphite portions of the 12 standard. And those contracts are set up to provide 13 on-call technical expertise to facilitate the staff's 14 review in drafting the NUREG and reg guide.

15 They were also used to answer staff 16 questions regarding the adequacy and use of Division 17 5, and they were used to provide the staff with the 18 technical basis and historical perspectives related to 19 Division 5.

20 So slide 37.

21 So this slide sort of provides an overview 22 of the philosophy we use for endorsement. As Jeff was 23 sort of alluding to, the rules in Division 5 have been 24 developed over the years. The NRC endorses ASME 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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73 Section III, Division 1, by incorporating it by 1

reference into 10 CFR 50.55(a).

2 Those rules apply to components that 3

operate at temperatures that are typically 700 degrees 4

Fahrenheit or less for carbon or carbon steels and 800 5

degrees Fahrenheit for -- or less for austenitic or 6

high-nickel alloys where creep effects are 7

insignificant.

8 In the

1970s, to facilitate the 9

construction of high-temperature reactors, ASME 10 developed five code cases that were intended to 11 replace or supplement in some casesSection III, 12 Division 1, and those are Code Cases 1592 through 13 1596.

14 And it was intended that these code cases 15 could be used as a guide with justification provided 16 by an applicant to supplement other Section III 17 subsections and appendices used to design components 18 operating at high temperatures. They were approved by 19 ASME in the '70s and endorsed by the staff in Reg 20 Guide 1.87 Revision 1.

21 ASME subsequently incorporated those five 22 code cases into Division 1 with the creation of ASME 23 Section III, Division 1, NH, and the NUREG uses these 24 code cases as a basis for the review of the 2017 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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74 edition of Division 5.

1 MEMBER HALNON: Hey, Jordan. This is Greg 2

Halnon. Just a quick question.

3 MR. HOELLMAN: Sure.

4 MEMBER HALNON: Since those code cases and 5

the review was done 45 years ago, did you do any 6

cursory look at it or a deeper look to make sure that 7

in today's standards and with the OE that we've 8

received over the last many reactor years that 9

everything is still good and able to stay with it in 10 this new review?

11 MR. HOELLMAN: Yeah. So we did do a 12 detailed historical review of the code cases, a 13 comparison between the code cases and what's in 14 Division 5 now, as well as a look at preliminary 15 safety evaluation reports that the staff developed.

16 We have also been -- the staff has been 17 involved in all of the working groups and subgroups on 18 the ASME code, and so we've been involved and aware 19 of, you know, the changes that have occurred. And so 20 we've looked at any differences and the improvements 21 that have been made over the years to the code. So it 22 was a detailed review of what was in the previous code 23 cases as well as the additional information.

24 I think we -- I'd say that we definitely 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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75 looked in more detail at what was added or changed 1

versus, you know, what remained the same.

2 MEMBER HALNON: And so you have high 3

confidence in the review back 45 years ago, is the 4

same that you would expect today moving forward on 5

materials and stuff?

6 MR. HOELLMAN: Right. Yeah.

7 MEMBER HALNON: Okay.

8 MR. HOELLMAN: And the code, you know, has 9

-- as it has been developed over the years, you know, 10 and incorporated into Division 1 in NH, the rules of 11 the code have, you know, incorporated the Division 1 12 standards that we have been endorsing via 10 CFR 13 50.55(a) over the years.

14 MR. HOELLMAN: Jeff, do you want to add 15 anything there?

16 MR. POEHLER: I just wanted to add that it 17 was within the scope of the contractor reviews to look 18 at whether the code case provisions were still 19 technically adequate.

20 MEMBER HALNON: That's what I was looking 21 for, to make sure that there is some -- that it just 22 wasn't --

23 MR. POEHLER: If that was their basis for 24 recommending something.

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76 MEMBER HALNON: Yeah.

1 MEMBER BALLINGER: This is Ron Ballinger.

2 I'm encouraged to see that 617 is going to be 3

incorporated into 1380. Was that originally the case?

4 MR. HOELLMAN: No. That -- well, we had 5

locations that weren't approved by ASME prior to the 6

initiation of our endorsement effort, and since 7

they've been incorporated and due to interest from, 8

you know, potential applicants, we have decided to 9

take on a separate activity to review those code 10 cases.

11 And because it sort of occurred, you know, 12 as we were getting to the end of our endorsement 13 review of Division 5, we have kind of decided that 14 let's continue with our current effort and take that 15 on in parallel.

16 And then I'll get to it later in the -- in 17 our next steps slide, but the plan currently is to, 18 you know, do the public comment period on our current 19 effort and incorporate it later and do another public 20 comment period, but limit it to the Alloy 617 code 21 cases.

22 MEMBER BALLINGER: So that will delay 1380 23 a little bit, though, right?

24 MR. HOELLMAN: It will delay the final 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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77 issuance of it maybe, but we're hoping that we can 1

sort of tackle that in parallel with the public 2

comment period and the final issuance of the reg 3

guide. But it will overlap a little bit, and that's 4

-- some schedule challenges will have to --

5 MEMBER BALLINGER: That's a very good 6

thing. I think -- we had a previous presentation 7

where we made a comment of why is 617 not included, 8

and the feedback that we got was that it was too early 9

because it had just been approved. But now it has 10 changed, and that's a very good thing, in my opinion.

11 MR. IYENGAR: This is Raj. May I 12 interrupt here, Jordan? Raj Iyangar. I just want to 13 tell you, Ron, we had talked about, discussed this 14 topic.

15 The code case was the -- 617 was passed, 16 approved late last year. So by then our Division 5 17 endorsement, the staff endorsement effort, had, you 18 know, been going on for a year and a half.

19 However, I think based on our discussion 20 we had, and based on the feedback we got from 21 industry, we had actually had a conflict with this in 22 a very agile way. I think Jeff and Jordan will talk 23 about it later. So that we don't delay the final 24 relief of the current -- the draft guide we are 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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78 proposing, but still incorporate the 617, and you know 1

the importance of that because it allows for such high 2

temperatures.

3 MEMBER BALLINGER: No, no, that's very 4

good. Thank you.

5 MR. HOELLMAN: Okay. So I'll continue a 6

little bit in describing how we approached the view.

7 So we compared the articles of ASME,Section III, 8

Division 5, HBB, which is the Class A metallic 9

pressure boundary components operating at elevated 10 temperature service. So we compared HBB to the 11 related areas of code cases 1592 through 1596 as an 12 approach to validate that the information present in 13 HBB is for high-temperature Class A components, which 14 is analogous to high-temperature Section III, Division 15 1, components addressed by the code cases.

16 The HBB provisions were reviewed with the 17 assumption that the components have safety-significant 18 functions similar to Division 1, Class 1, components.

19 In sort of the same manner, we compared 20 the HCB rules, which is the Class B metallic 21 components at elevated temperature service, to ASME 22 code NC and HBB since HCB, which is Class B again, is 23 for high-temperature Class B components, analogous to 24 Class 2 components, and NC, but operate at high 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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79 temperatures like the components addressed by HBB.

1 So this is where it gets a little 2

complicated, and Jeff's little magic decoder table 3

comes in handy.

4 So HCB provisions were reviewed with the 5

assumption that the components have similar functions 6

to Division 1, Class 2, components. When we get to 7

HGB, which is the core support structures, we compared 8

or the code sort of compares them to HBB, because core 9

support structures operate at the same 10 high-temperature range as that established for the 11 Class A components under HBB.

12 When evaluating the provisions of HAA and 13 HAB, which is the general requirements, HAA is for 14 metallic materials and HAB is for graphite materials.

15 We compared these to the 2017 edition of Section III 16 NCA, which the staff endorsed in 50.55(a).

17 When using -- so one of the limitations or 18 exceptions we're proposing is consistent with Section 19 III, Division 1. Where Division 5 references Division 20 1, applicants or licensees should follow any of the 21 applicable conditions for Division 1 that are 22 identified in 50.55(a).

23 I hope I didn't confuse that too much. So 24 we can move on to the next slide, if that's okay.

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80 So this slide just details the contractor 1

assignments and provides links to the specific 2

contractor reports that were used in combination with 3

the NRC staff's independent technical expertise to 4

develop the technical basis for the findings in 5

NUREG-2245.

6 As I mentioned before, the way we assign 7

the -- well, the way that contractor assignments are 8

set up does have some overlap, so we did ensure that 9

we were scheduling coordination meetings between the 10 different contractors and setting up that SharePoint 11 site where we could all collaborate, because some of 12 the rules, for example, in the 3000 reference, the 13 rules in the 2000s portions of the code.

14 And so some of the recommendations 15 provided by the contractors in 3000 relied on some of 16 the findings in -- or the recommendations in 2000 that 17

-- you know, for example, PNNL was not reviewing the 18 2000 portions of the code, and so we needed to make 19 sure that we were all coordinated and could resolve 20 issues between the different contractors.

21 So we can move on to slide 39, and I'm 22 going to turn it back over to Jeff to walk through 23 some of the expected limitations and exceptions we are 24 proposing throughout our review.

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81 MR. POEHLER: Thanks. Thanks, Jordan. So 1

I'm going to talk a little bit more about the review 2

process for general requirements. Jordan touched on 3

that already.

4 But the process -- basically, the staff 5

compared the 2017 edition of Division 5 HAA and HAB to 6

the 2017 edition of the ASME code,Section III, NCA, 7

to ensure consistency with what the NRC has endorsed 8

in 10 CFR 50.55(a), or I guess I should say 9

incorporated by reference.

10 Similarly, the staff compared the 2017 11 edition of Division 5, HAA and HAB, to the 2019 12 edition of the ASME code, Division 5, HAA and HAB, to 13 ensure consistency with those items that were 14 corrected in the 2019 edition.

15 Just a little more background on that, the 16 NRC does participate in the relevant code committees 17 related to general requirements, and the staff 18 recognized that some changes in the 2019 edition of 19 Section III NCA were needed and were not captured in 20 the 2017 edition of HAA and HAB.

21 The staff, therefore, also identified 22 exceptions and limitations when there were differences 23 between the 2017 and 2019 editions of Division 5, HAA 24 and HAB. Even though the rulemaking to IBR are 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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82 incorporated by reference, the 2019 edition of HCA 1

into 10 CFR 50.55(a) is not quite final.

2 So now I'm going to give a couple of 3

examples of exceptions and limitations related to 4

general requirements. The first one is related to a 5

change in ASME Section III NCA to allow certifying 6

engineers who are not registered professional 7

engineers.

8 The staff conditioned this in its 9

rulemaking to incorporate by reference the 2017 10 edition of Section III NCA to require the certifying 11 engineers also to be a registered professional 12 engineer. Therefore, the limitation in the draft 13 guide is for consistency with the condition in 10 CFR 14 50.55(a).

15 The second limitation is related to 16 standards used for accreditation of providers of 17 calibration and testing services.

The ILAC 18 accreditation process relies on the ISO/IEC 17025 19 standard, and use of the 2005 edition of ISO/IEC 17025 20 was endorsed by the NRC through an SER with several 21 conditions.

22 In 2017, ISO issued the 2017 edition of 23 ISO/IEC 17025, which the NRC endorsed again through 24 another SER with some additional conditions.

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83 Therefore, the NRC is proposing a limitation to 1

Division 5 to make it consistent with the NRC's latest 2

SER.

3 Next slide, please.

4 Now I am going to talk about some of the 5

exceptions and limitations that the staff is proposing 6

in the area of mechanical design, and these were 7

identified for several reasons. One of those is 8

consistency with Section III, Division 1, conditions 9

in 10 CFR 50.55(a).

10 An example of that is the condition on 11 socket weld design, and that condition requires a 12 larger leg length on socket welds than Section III, 13 Division 1, allows.

14 And a second reason would be consistency 15 with Reg Guide 1.87 conditions on Code Case 1592. One 16 example of that is the use of strain-controlled 17 buckling factors, and this limitation is based on a 18 limitation in Reg Guide 1.87 on Code Case 1592 related 19 to the situation where you could have elastic follow 20 up occurring.

21 And another reason that we identified 22 condition -- or limitations and exceptions is a lack 23 of guidance in Section III, Division 5. And some 24 examples of that are for inelastic analysis for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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84 meeting the HBB-T deformation limits.

1 And as I discussed earlier, there are no 2

material models for inelastic analysis currently in 3

the code. So the staff would want to review any such 4

models that were proposed for use by applicants.

5 Another example is related to stress 6

relaxation cracking, and I am going to talk about that 7

on the next slide.

8 So let's go to the next slide, please.

9 So the limitation here is when using 10 HBB-T-1710, applicants and licensees should develop 11 their own plans to address the potential for stress 12 relaxation cracking in their designs. The basis for 13 this is that stress relaxation cracking is a mechanism 14 causing enhanced creep crack growth in certain 15 materials caused by relaxation of weld residual 16 stresses in components in high-temperature service.

17 Section III, Division 5, does not contain 18 any provisions addressing stress relaxation cracking.

19 And also, there is a lot of literature on stress 20 relaxation cracking, and there are approaches that can 21 be used to address it that could be used by applicants 22 but they are not in the code. So that's why we 23 included a limitation for applicants to, you know, 24 explain how they are addressing this phenomenon.

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85 Next slide?

1 So this slide discusses the review process 2

for metallic and graphitic materials. So unlike in 3

the area of general requirements and mechanical 4

design. In these areas, the staff did not primarily 5

rely on previous reviews of the code cases.

6 For metallic materials, the contractor 7

performed independent analysis of materials properties 8

and allowable stresses. The staff also received 9

additional input by subject matter experts familiar 10 with the development of Section III, Division 5, in 11 the area of materials properties. And we also 12 considered that an input.

13 With respect to graphite provisions, they 14 weren't in any previous code cases. They were new to 15 Division 5. Therefore, the staff contracted for 16 technical review of the graphite portions of Division 17 5 by subject matter experts.

18 I am going to discuss the review of both 19 metallic and graphitic materials in more detail in 20 subsequent slides.

21 So next slide, please.

22 So with respect to metallic materials 23 properties, so -- is there a question? Sorry.

24 Okay.

In some

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86 independent analysis determined properties and 1

allowable stresses with lower values than the code, 2

suggesting that code values are non-conservative. And 3

those were -- contractor reports were primarily by Oak 4

Ridge National Laboratory, which covered allowable 5

stresses in Mandatory Appendix HBB-I-14 and also some 6

other material properties in that appendix.

7 Oak Ridge performed independent analysis 8

of the metallic materials properties and allowable 9

stresses, and that analysis was based on available 10 data from a literature search, and it used the stated 11 criteria for determining, you know, allowable stresses 12 in Section III, Division 5.

13 Methodology used was ASME standard 14 practice as far as that can be defined.

15 There is a report by Numark that found --

16 that looked at the isochronous stress strain curves 17 and suggested some of those could be non-conservative.

18 We had Argonne National Laboratory assist 19 with the review of weld strength reduction factors, 20 which were found to not be non-conservative.

21 So lower values of allowable stresses were 22 typically only at higher temperatures and longer times 23 for the time-dependent properties. The NRC staff did 24 consider these findings in a holistic manner, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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87 including how these properties are used, inherent 1

conservatisms in the Division 5 design rules, and 2

historical context. And input from ANL provided 3

historical context and perspective on materials 4

properties.

5 Next slide, please?

6 So for metallic materials, limitations are 7

typically in the form of a maximum temperature limit 8

that is more restrictive than allowed by Division 5.

9 These limitations are typically on the time-dependent 10 allowable stresses. The table here shows these limits 11 for the materials where those apply, and you can see 12 that the materials involved here, the properties were 13 typically -- the SMT, which is -- can be controlled by 14 the time-dependent allowable stress, the S sub T, 15 which is the time-dependent allowable stress, and the 16 S sub R, which is the stress to rupture.

17 For non-chrome 1 Moly-Vanadium, we took a 18 different approach. The 2019 Section III, Division 5, 19 properties were endorsed in lieu of the 2017 Section 20 III, Division 5, properties. And that was done 21 because ASME updated, in the 2019 edition, the values 22 for non-chrome. And those compared well with the 23 independent analysis thoughts, while the 2017 values 24 in Division 5 appeared to be somewhat non-conservative 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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88 compared to the independent analysis.

1 Next slide, please.

2 So I'm going to talk a little more about 3

the basis about how the limitation of one type of 4

allowable stress was determined. The example here is 5

Type 304 stainless steel where the independent 6

analysis suggested significant non-conservatism of the 7

Section III, Division 5, S sub T values for most times 8

and temperatures.

9 At 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, non-conservatism was 10 suggested at temperature -- any temperature greater 11 than 850 degrees F or 450 degrees C, but depending on 12 whether you are in the U.S. customary table or the SI 13 table.

14 This is based on independent analysis 15 values more than 10 percent lower than the Section 16 III, Division 5, values. Most of the apparent 17 non-conservatism here was driven by the tertiary creep 18 criterion for S sub T.

19 And the use of the time to tertiary creep 20 as one of the three criteria for time-independent 21 allowable stresses is problematic. There is less data 22 for tertiary creep than for creep rupture in general.

23 It's a smaller database. It is often difficult to 24 identify the onset of tertiary creep.

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89 And in materials that demonstrate 1

non-classical creep behavior the onset of tertiary 2

creep can be relatively early, which results in lower 3

times the tertiary creep and a slower -- lower 4

allowable stresses.

5 So the ASME code has been deliberating 6

modification or elimination of the tertiary creep 7

criterion. I mean, they haven't done it yet.

8 There is a proposal to revise the 9

allowable stress for Type 304 and Type 316 to be made 10 in the ASME code committees, and it will use a linear 11 multiplier on the rupture time to estimate the time of 12 tertiary creep, which will increase the number of 13 tertiary creep data points.

14 So this issue for Type 304 was mitigated 15 by ANL performing an alternate analysis using a 16 different approach for tertiary creep data. And this 17 analysis showed significant non-conservatism only at 18 temperatures greater than 1,300 -- or greater than or 19 equal to 1,300 degrees Fahrenheit or 700 degrees C.

20 So next slide, please.

21 Okay. Now moving on to discussing the 22 review of graphite materials and design, so Numark 23 Associates provided a

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90 Structures, Subpart A, Graphite Materials.

1 The staff completed the review of the 2

above report and all applicable sections of Section 3

III, Division 5, and obtained clarifications and 4

feedback from NRC contractors, including Numark and 5

Idaho National Laboratory, in order to come up with 6

the conclusions identified in the NUREG.

7 The staff's independent review of the code 8

requirements considered the holistic design of 9

graphite core support structures.

10 Next slide, please.

11 So I am going to talk a little more about 12 some of the exceptions and limitations the staff is 13 proposing for graphite. So for graphic materials and 14

designs, several of the limitations can be 15 characterized as situations where Division 5 has a 16 numerical parameter limit, but the staff is not 17 convinced the limit is generally applicable to all 18 designs.

19 And so design-specific justification is 20 requested for the parameter value in these cases as a 21 limitation. And this table shows the provisions where 22 the staff identified such limitations, including the 23 parameter affected in the Division 5 limit, and those 24 include weight loss limit, cohesive life limit, gas 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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91 flow velocity, allowed repair depth. So that was a 1

common theme for several of the limitations.

2 Next slide?

3 MEMBER KIRCHNER: Could you -- this is 4

Walt Kirchner. Can you provide a little more detail 5

on the first one, oxidation? That seems like a 6

substantial amount of oxidation weight loss.

7 MR. POEHLER: It does. For that I would 8

call on -- if we have either Matthew Gordon or Steve 9

Downey on the line? Or, if not, I would -- I would 10 ask Will Windes if he can chime in on that, if he is 11 still on.

12 DR. WINDES: Will is here, but I -- if 13 somebody else from the NRC wants to talk about it 14 first, that would be perfect.

15 MR. POEHLER: Yeah. I mean, Will was not, 16 you know, directly involved with the condition.

17 MEMBER KIRCHNER: Yeah. It just strikes 18 me as -- boy, that strikes me as a large oxidation 19 loss.

20 DR. WINDES: Yeah.

21 MEMBER KIRCHNER: So way beyond anything 22 a designer would probably want to incorporate in an 23 actual operating envelope.

24 DR. WINDES: Right. So here is the --

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92 here is the issue. And you're absolutely correct.

1 This oxidation -- the oxidation limits in the code are 2

being changed rapidly and dramatically as we speak.

3 We have a task group that is working on coming up with 4

something that is much more -- makes much more sense 5

and is much more usable.

6 There has been a number of papers written 7

in the last year, so that kind of talks about not so 8

much what is the weight loss, but what is the effect 9

of weight loss.

10 So 30 percent -- and then, of course, the 11 real issue -- and I think this is -- and I'm 12 speculating now. I think that one of the main issues 13 that the NRC had was, where is the weight loss 14 occurring?

15 So if it's occurring in the material --

16 excuse me, the structural graphite that is directly 17 surrounding the fuel, this could be extraordinarily 18 significant. If this is something that's occurring --

19 these limits are occurring in something in the core 20 support structures, again, very significant.

21 If it's occurring in the outer reflector 22 blocks, which are just basically outside of the core 23 area, then it may not be as catastrophic. It's 24 obviously going to be beyond what any kind of designer 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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93 wants. And so I think you're absolutely correct.

1 This number is there.

2 The other number -- I want to call your 3

attention also to number 2, which is something that is 4

in hot debate. This is what Dave Petti alluded to 5

earlier this morning where this is plus 10 percent 6

over the crossover line for the dimensional change.

7 And this is something that no -- this is 8

an area that nobody has ever operated their reactors 9

in. And I think that the NRC is quite correct in 10 identifying this one as a problem as well.

11 So these are very hot topics that we are 12 changing right now.

13 MEMBER KIRCHNER: Yeah. I just can't 14 imagine, with numbers 1 and 2 there, going anywhere 15 near that in an actual design. Wow.

16 CHAIR RICCARDELLA: No. That's why this 17 table is requesting design-specific justification for 18 these limits, if they use them.

19 MEMBER KIRCHNER: Yeah. It could be a lot 20 less for certain locations that could be tolerable, 21 and then it --

22 DR. WINDES: It could be more for certain 23 locations, because I believe at 30 percent the code 24 says above that you take out -- you just consider the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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94 component has no strength basically. None.

1 MEMBER KIRCHNER: Yeah. That's what 2

you're -- with that amount of oxidation, you probably 3

have no structural rigidity or strength left in the 4

component. Going through that bend in the curve, and 5

then going plus 10 percent on the life limit, my 6

goodness, like I alluded to earlier, in the N reactor, 7

they had issues like that. And it's just -- it's not 8

practicable for an actual reactor design for a number 9

of reasons, not to get into here.

10 DR. WINDES: Right. And I can sum it up.

11 It's too much risk. You can operate a reactor --

12 obviously, they have -- and reactors are a great 13 example of it. But, quite frankly, for a civilian 14 reactor, it's just too much risk. We don't know 15 what's going to happen above crossover.

16 And there is just so little data, and you 17 just cannot go in and really predict what is going to 18 happen. So, again, these are -- these are things that 19 even before the NRC tagged these as hot button topics 20 we were already working on them, because we ourselves 21 have identified these as real gaps in the code, and 22 significant ones that need to be addressed sooner 23 rather than later for the licensee applicants if they 24 are going to use the code.

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95 MEMBER REMPE: So what's motivating folks 1

to want to even go to 30 percent? Because actually 2

weren't the AGRs -- their limits are actually higher 3

for oxidation. But in the U.S., has that been allowed 4

for things like Fort Saint Vrain?

5 DR. WINDES: Well, yeah. See, and this is 6

where -- Joy, boy, you always put your finger right on 7

the issues. Yeah. The real problem is is that the 8

AGRs in the U.K. have technically suffered much 9

greater weight losses, and they are perfectly safe in 10 operations.

11 So the -- and then, in the United States, 12 it has been much, much more extremely conservative.

13 You know, nothing more than, say, 10 percent weight 14 loss. But, of course, they don't tell you where that 15 10 percent weight loss occurs in the code as it exists 16 now, which is, again, an issue we're talking about --

17 or fixing.

18 So what we tried to do was come up with a 19 happy medium where we said if you go in -- and 20 anything up to 30 percent, you need to justify with 21 your design that this is okay. But we're just not 22 even going to consider anything over 30 percent. We 23 just can't.

24 Even though there is examples of other 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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96 designs -- namely, the AGRs -- suffering oxidation 1

beyond 30 percent, without any safety consequences, we 2

just don't want to take that risk on. We're going to 3

limit it at 30 percent, and you -- if you get close to 4

30 percent, boy, you'd better have a really good 5

justification for using it.

6 Anything -- and, of course, less and less, 7

you don't have to have as much effort to go in and 8

show that everything is going to be safe. So if 9

you're like one or two percent oxidation, then that's 10 not nearly as onerous as, say, 29 or 30 percent. Do 11 you understand what I'm saying?

12 MEMBER REMPE: Yeah. But I don't think 13 you're understanding my question.

14 DR. WINDES: Ah. Sorry.

15 MEMBER REMPE: Why is it proposed to go 16 from 10 percent to 30 percent? Are there some design 17 developers out there that are saying we think we need 18 to go much higher because our design is going to be 19 approaching 30 percent?

20 DR. WINDES: No. What we were trying to 21 do -- and, again, please forgive us because we were 22 basically designing in a vacuum. There has been no 23 previous designs. What we were trying to do is make 24 it as universally applicable to as many and all 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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97 designs as we could.

1 Just like we don't limit it to one grade 2

of graphite. If you want to use a different grade 3

than somebody else, you are allowed to do that. But 4

you have to do certain things to justify the use of 5

using a graphite or going to those kinds of high 6

oxidation.

7 So it was -- it was basically an attempt 8

to go in and have -- accommodate as many designs as 9

possible.

10 MEMBER REMPE: Okay. Thank you.

11 DR. WINDES: Sure.

12 MEMBER KIRCHNER: I think, too, the 13 distinction with regard to the AGR is that, if I 14 remember right, these are pressure tube reactors. The 15 graphite is not serving a structural function. It is 16 there to be a moderator.

17 So what happens to the graphite in an AGR 18 is not a good example for, say, a pebble bed or a 19 modular HGGR design. Completely different design 20 construct.

21 MEMBER REMPE: That's exactly why I was 22 asking is why go so much higher, because I'm not 23 aware, but I don't know of all the designs that are 24 being proposed and what they are thinking of. But I 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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98 wasn't aware that they would need to go so much 1

higher.

2 DR. WINDES: Right. And, Joy, I haven't 3

-- I mean, I can only think of one design, and it's 4

way out there. I'm not even sure it is being fully 5

funded at this point. It's just an idea. But nobody 6

-- nobody designs I think their reactors for 30 7

percent or more oxidation. That would be sort of 8

effectively operating in air with graphite at higher 9

temperatures. So that's kind of crazy.

10 And so at that point you're right. I just 11 don't -- I'm not aware of any, but we didn't want to 12 limit. The code is there to try to be as universally 13 applicable as possible. We didn't want to limit 14 anybody. And because there is -- there are designs 15 out there that can operate at the higher oxidation, we 16 wanted to make sure that they -- we had a higher than, 17 say, just a very conservative five to 10 percent mass 18 loss. Okay?

19 CHAIR RICCARDELLA: Okay. So we're 20 reached the published time at -- for the meeting to 21 end. We've got about four or five more slides. I'm 22 going to propose that we continue on and finish.

23 Hopefully we finish in 15 minutes or so.

24 MR. POEHLER: Yeah. These should go 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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99 pretty quick, I think, probably famous last words, but 1

-- anyway, so, yeah, this -- so this slide discusses 2

an exception or limitation that doesn't fit the mold 3

of the ones I discussed from the previous slide.

4 This has to do with a provision in 5

HHA-3330 that says you have to design to allow for 6

in-service inspection. But, if necessary, you can 7

replace in-service inspection by operational 8

monitoring. And we are not endorsing -- the staff is 9

not proposing to endorse this provision because 10 requirements for in-service inspection are outside of 11 the scope of Section III, Division 5, HHA.

12 And the provision related to operational 13 monitoring is the one that the staff finds to be out 14 of scope. So that's why we proposed the limitation to 15 not endorse HHA-3330(g).

16 Let's go to the next slide.

17 Okay. Shifting gears a little here, so 18 this slide just talks about quality group 19 classifications. Those are covered in Appendix A of 20 DG-1380, and that provides the staff's guidance on 21 quality group classifications. And the approach is 22 very similar to that in NEI-1804.

23 Quality Group A is safety-related systems, 24 structures, and components. For that, you can use --

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100 use that for components that have safety-related -- or 1

safety-related systems, structures, and components 2

that have safety significance.

3 Quality Group B is for safety-related 4

systems, structures, and components, but Class B 5

components in Division 5 that have low safety 6

significance.

7 Quality Group C is for non-safety-related 8

systems, structures, and components, with no special 9

treatment, or, I'm sorry, non-safety-related systems, 10 structures, and components with safety significance.

11 And for that you would use ASME Section VIII, Division 12 1 or 2 rules.

13 And then Quality Group D

is for 14 non-safety-related systems, structures, and components 15 with no special treatment, and the owner would 16 establish standards for use for those. And Quality 17 Group D can also be described as systems, structures, 18 and components having low safety significance or no 19 safety significance. I think it's no safety 20 significance.

21 CHAIR RICCARDELLA: So, Jeff, does Section 22 VIII, Div 1 or 2 have high-temperature considerations 23 in them?

24 MR. POEHLER: Section VIII does, yeah.

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101 CHAIR RICCARDELLA: All right.

1 MR. POEHLER: Or pressure vessel. So --

2 CHAIR RICCARDELLA: Just another question 3

for -- I'm just curious as to -- for Division 5, why 4

they got away from Class 1 and Class 2 and then call 5

them Class A and Class B? That's kind of curious to 6

me. You know, everyone has gotten familiar with the 7

concept of a Class 1 component.

8 MR. POEHLER: Yeah. That sounds like --

9 I would probably call on Sam Sham to chime in on that 10 because I really don't know. Are you there, Sam?

11 DR. SHAM: Yes. It was just a distinction 12 that -- when the group puts together Division 5, to 13 distinguish between the rules for the high temperature 14 and the ones in Division 1.

15 CHAIR RICCARDELLA: Okay.

16 MR. POEHLER: Thanks, Sam.

17 CHAIR RICCARDELLA: Thank you.

18 MR. POEHLER: Okay. Next slide, please.

19 MR. HOELLMAN: All right. Jeff, I think 20 this is me again.

21 MR. POEHLER: All right. Thanks, Jordan.

22 MR. HOELLMAN: Yep. So this sort of just 23 summarizes what I talked about earlier, and I'll try 24 to move through it quickly, because I know we're 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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102 running out of time.

1 So we completed our technical review of 2

the 2017 edition, and we're in the process of 3

finalizing the documents for public comment.

4 NUREG-2245 provides the technical basis for the staff 5

positions in DG-1380, which is a proposed revision to 6

Reg Guide 1.87.

7 Jeff just discussed some examples of the 8

exceptions and limitations we expect to include in the 9

draft guide, so I won't spend much more time on that.

10 So let's move to 52, and this just discusses our next 11 steps. So we're going to finalize the documents for 12 public comment.

13 We'll address public comments and make any 14 changes necessary in parallel with our effort to 15 review for endorsement the Alloy 617 code cases. We 16 will plan to -- our current plan is to supplement the 17 draft guide with the Alloy 617 code cases, and any 18 limitations or exceptions we think are necessary 19 there, issue that for a separate public comment 20 period, limited to only the Alloy 617 code cases, and 21 then issue the final reg guide, likely in the early 22 2022 timeframe.

23 CHAIR RICCARDELLA: So when do you 24 anticipate that the draft -- the original draft guide 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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103 will go out for public comment?

1 MR. HOELLMAN: Well, we're shooting for 2

the end of this month.

3 CHAIR RICCARDELLA: Okay.

4 MR. HOELLMAN: It does take -- I'm 5

realizing that it takes a little bit more time to 6

issue a NUREG than some other documents. So there's 7

a little bit of process period there, but we're close.

8 And like I said, the technical review is done. It's 9

just, you know, working through the internal reviews, 10 and whatnot, to get the things out the door for public 11 comment. And it will be a 60-day public comment 12 period.

13 CHAIR RICCARDELLA: Okay. Okay. Well, I 14 thank the staff for an excellent presentation. Very 15 informative.

16 And I want -- at this point, I'll go 17 around, see if any of the members have any additional 18 comments or questions. I hear silence.

19 MEMBER REMPE: Pete, this is Joy.

20 CHAIR RICCARDELLA: Yeah.

21 MEMBER REMPE: Is NUREG-2245 available for 22 public -- to the public, or what's the status of that 23 document?

24 MR. HOELLMAN: This is Jordan. Both of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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104 the documents are being finalized internally now. So 1

nothing is publicly available yet. We are shooting 2

for the end of this month to get things out the door 3

and publicly available. We wanted to have everything 4

publicly available before this briefing, but we just 5

didn't quite make it. So apologize for that.

6 MEMBER REMPE: But when it is available, 7

please provide a copy to Kent, so he can -- of each 8

document for us, please.

9 MR. HOELLMAN: Definitely. Yep. Thanks, 10 Joy.

11 CHAIR RICCARDELLA: Thank you. Any other 12 member comments or questions? Okay.

13 So then, at this point, we'll turn to the 14 public and see if there are any public comments. Can 15 someone confirm the bridge line is open?

16 MR. DASHIELL: The public bridge line is 17 open for comments.

18 CHAIR RICCARDELLA: Okay. So if there is 19 anybody from the public out there that would like to 20 make a comment, please state your name and make your 21 comment. Hearing none -- I'm sorry. Go ahead.

22 MS. BOUDART: Could I ask a question? I'm 23 a member of the public.

24 CHAIR RICCARDELLA: Yeah. You can make a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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105 comment, yeah.

1 MS. BOUDART: I can make a comment. Okay.

2 Well, I didn't want to make a comment. I wanted -- I 3

was so fascinated by the discussion of graphite, and 4

of course everybody was. We really -- the discussion 5

really got kind of stuck there.

6 I am very interested in the explosion at 7

Chernobyl and the fact that graphite was considered a 8

moderator of the neutron flux there. And that when 9

the negative coefficient was reached, I guess that the 10 liquid moderator turned to bubbles, so that the 11 neutron flux was full force on the graphite and it 12 couldn't handle it.

13 I'm wondering if somebody could -- if 14 there is any comment on the quality of the graphite, 15 because we went -- you went into so much detail about 16 the quality of the graphite and how important that is.

17 Do you -- does anybody think that a different quality 18 of graphite could have prevented that explosion?

19 MR. MOORE: This is Scott Moore for the 20 ACRS. Could the member of the public please state 21 your name for the record.

22 MS. BOUDART: Oh, I'm sorry. I'm Jan 23 Boudart, and I'm a board member of the Nuclear Energy 24 Information Service.

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106 MR. MOORE: Thank you, Ms. Boudart.

1 CHAIR RICCARDELLA: Does anybody -- we 2

normally don't answer questions, but does anybody wish 3

to make a comment that knows more than I do about what 4

happened at Chernobyl? I think -- I think --

5 MEMBER REMPE: Pete, this is Joy. There 6

is a bit of confusion in the comment that was provided 7

with respect to what this --

8 MS. BOUDART: Yeah. I do --

9 MEMBER REMPE: -- what is the moderator.

10 And, again, we don't respond to public comments at 11 this meeting, but I strongly suggest that the member 12 of the public obtain a general overview article about 13 the Chernobyl reactor design.

14 CHAIR RICCARDELLA: Yeah. Perhaps you 15 could send that question to Kent Howard, the public 16 official -- the government official for the meeting, 17 and he could maybe coordinate a reply.

18 MS. BOUDART: Okay. I appreciate it.

19 Thank you.

20 CHAIR RICCARDELLA: Any other members of 21 the public?

22 MS. WALKER: Yeah. Can you hear me?

23 CHAIR RICCARDELLA: Yes.

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107 of the public. I was wondering if the ASME code cases 1

can be made public. We're not able to access the code 2

cases.

3 CHAIR RICCARDELLA: Well, they are 4

available for purchase. The codes generally are 5

available for purchase by ASME. It's an ASME product.

6 MS. WALKER: Right. But we thought, you 7

know, it's a standard, so I was just wondering -- so 8

you have to -- you have to pay to know the actual 9

standard.

10 MR. HOELLMAN: This is Jordan.

11 MS. WALKER: That won't be made available.

12 MR. HOELLMAN: This is Jordan Hoellman.

13 I think when we release the documents for public 14 comment there are instructions on how you can obtain 15 a copy of the code. I think the public document room 16 does have a copy for public inspection during public 17 comment periods, but that will all be included in the 18 Federal Register Notice issuing the documents for 19 public comment.

20 MS. WALKER: I have a -- I'm particularly 21 interested in the code case regarding the inspection 22 of the nuclear pressure vessels for storage. And 23 there was a code case that was just published, right, 24 as an approval of an inspection/maintenance program 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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108 was approved for the canisters at San Onofre 1

specifically.

2 But anyway, I wasn't able to get the 3

current code case. And I did purchase one, but I 4

don't think it listed the current one. So maybe I 5

could email somebody there, just to verify whether 6

what was published that I purchased is the most 7

current. Would somebody be willing to do that?

8 CHAIR RICCARDELLA: Yeah. That --

9 MS. WALKER: It's a little bit -- it's a 10 little bit challenging being a member of the public 11 and these code cases being referenced, but we can't 12 access them.

13 CHAIR RICCARDELLA: That code case is 14 totally separate from this meeting, which is on 15 high-temperature code cases.

16 MS. WALKER: I understand.

17 CHAIR RICCARDELLA: There's probably --

18 MS. WALKER: I understand, but it's about 19 20 CHAIR RICCARDELLA: You know, it's 21 probably more appropriate to contact someone from ASME 22 about whether that's the most current code, not the 23 NRC.

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109 getting through to them or getting the information, so 1

-- anyway, thank you.

2 CHAIR RICCARDELLA: Okay. Any other 3

members of the public that would like to make a 4

comment?

5 Okay. With that, I will close the 6

meeting, and I thank everybody for their participation 7

and all. And for the members, we'll see you shortly 8

for the meeting on probabilistic fracture mechanics 9

this afternoon.

10 (Whereupon, the above-entitled matter went 11 off the record at 12:15 p.m.)

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Overview of Section III, Division 5 Advisory Committee for Reactor Safeguards July 20, 2021 Jeff Poehler, Sr. Materials Engineer Reactor Engineering Branch Office of Nuclear Regulatory Research 1

ASME Section III, Rules for Construction of Nuclear Facility Components -

Division 5, High Temperature Reactors ASME Section III Division 5 Scope

- Division 5 rules govern the construction of vessels, piping, pumps, valves, supports, core support structures and nonmetallic core components for use in high temperature reactor systems and their supporting systems o Construction, as used here, is an all-inclusive term that includes material, design, fabrication, installation, examination, testing, overpressure protection, inspection, stamping, and certification

- High temperature reactors include

• Gas-cooled reactors (HTGR, VHTR, GFR)

• Liquid metal reactors (SFR, LFR)

• Molten salt reactors, liquid fuel (MSR) or solid fuel (FHR) 2

Examples of Different Advanced Reactor Designs Being Developed By Industry Terrestrial Energy IMSR Fast Reactors Molten Salt Reactors Elysium, MCSFR TerraPower MCFR Gas Reactors Flibe Energy LFTR (thorium)

X-Energy, Xe-100 Framatome SC-HTGR General Atomic EM2 (Gas-cooled Fast Reactor)

Ultra Safe Nuclear MMR Westinghouse eVinci Heat Pipe Reactor GE Hitachi PRISM TerraPower, TWR Advanced Reactor Concepts, ARC-100 Westinghouse, LFR Oklo, Aurora TerraPower & GEH Natrium Kairos Power KP-FHR ThorCon Moltex Energy, SSR 3

Division 5 - A Component Code

• Division 5 is organized by Code Classes:

- Class A, Class B, Class SM for metallic components -

• Class A is analogous to Class 1 in Section III, Division 1

• Class B is analogous to Class 2 in Section III, Division 1

• Class SM is for metallic core supports

- Class SN for non-metallic components - e.g. graphite core supports

• Division 5 recognizes the different levels of importance associated with the function of each component as related to the safe operation of the advanced reactor plant

• The Code Classes allow a choice of rules that provide a reasonable assurance of structural integrity and quality commensurate with the relative importance assigned to the individual components of the advanced reactor plant 4

Section III, Division 5 Rules for Metallic Components do not address

• Deterioration in service due to

- Corrosion

- Mass transfer phenomena

- Radiation effects

- Other material instabilities

• Continued functional performance of deformation-sensitive structures such as valves and pumps 5

History of Construction Rules for High Temperature Reactor Components

• 159X Code Cases - complete construction rules for elevated temperature pressure boundary metallic components in early 1970s

• Regulatory Guide 1.87 (Rev 1, June 1975) endorsed 159X Code Cases with conditions

• Code Case series 1592-1596 converted to Code Case N-47, which later formed the basis for Section III, Division 1, Subsection NH

• Division 5 first published in 2011, combined NH, other high-temperature code cases, and rules for graphite core components (new).

6

Class Subsection Subpart Subsection ID Title Scope Class A, B, & SM A

HAA Metallic Materials Metallic Class SN B

HAB Graphite and Composite Materials Nonmetallic Class A A

HBA Low Temperature Service Metallic Class A B

HBB Elevated Temperature Service Metallic Class B A

HCA Low Temperature Service Metallic Class B B

HCB Elevated Temperature Service Metallic Class A & B HF A

HFA Low Temperature Service Metallic Class SM A

HGA Low Temperature Service Metallic Class SM B

HGB Elevated Temperature Service Metallic Class SN A

HHA Graphite Materials Graphite Class SN B

HHB Composite Materials Composite HH Class B Metallic Pressure Boundary Components General Requirements Class A Metallic Pressure Boundary Components Class A and Class B Metallic Supports Class SN Nonmetallic Core Components Class SM Metallic Core Support Structures HA HB HC HG Section III Division 5 Organization 7

Temperature Boundaries for Class A Components Maximum Use Temperature Metal Temperature Design Lifetime No Creep Effects Creep Does Not Affect Cyclic Life (Negligible Creep Regime)

Creep Affects Cyclic Life (Creep-fatigue Interaction)

Division 5 Division 1 Negligible Creep Temperature Code Temperature Boundary (700F ferritic; 800F austenitic) 8

HBB Materials and Design Data Limited set of materials:

- Type 304 Stainless Steel*

- Type 316 Stainless Steel*

- Alloy 800H

- 2.25Cr-1Mo

- 9Cr-1Mo-V (Grade 91)

- Alloy 617 (Code Cases N-872 and N-898)

Design parameters are mostly self contained in Division 5, except the following contained in Section II:

- Elastic constants

- Thermal properties

- Part of yield strength () table

- Part of ultimate tensile strength () table Minimum carbon content of 0.04 weight % required for better high temperature properties - Type 304H and Type 316H - this designation is not used in Section III-5.

9

Failure Modes Addressed by Section III-5 Failure Mode Type Prevented By Location Analysis Method(s)

Plastic collapse Load controlled Primary load design HBB-3000 Elastic Creep-rupture Load controlled Primary load design HBB-3000 Elastic Creep-fatigue Deformation controlled Creep-fatigue rules HBB-T Elastic, Inelastic, EPP Gross distortion due to incremental collapse and ratcheting Deformation controlled Strain limits HBB-T Elastic, Inelastic, EPP Buckling due to short-term loadings Load controlled or strain controlled, or both Buckling limits (time-independent)

HBB-T Elastic, Inelastic Creep buckling due to long term loadings Load controlled or strain controlled, or both Buckling limits (time-dependent)

HBB-T Elastic, Inelastic 10

HBB Primary Load Design

• Based on elastic analysis.

• Load-controlled

• Uses stress classification and linearization.

• Design and service level load checks.

• Accounts for thermal aging effects with factors on yield and ultimate strength

• Welds: Strength reduction factor applied Single temperature, pressure, and set of forces Time-independent Uses allowable stress

Very similar to Section I and VIII Time-history of loading Time-dependent Uses the allowable stress Unique to Division 5 Design Load Service Load 11

HBB - Allowable Stresses

• Both time-dependent and time-independent allowable stresses included.

• S0 - Allowable stress for design loadings

• Service Level Loading Allowable stresses

- Sm - Time independent

- St - Time dependent

- Smt - Allowable limit for general primary membrane stress for Service Level A and B

- Sr - Expected minimum stress-to-rupture. Used for Level D limits and in deformation-controlled analyses (HBB-T) 12

HBB - Basis for Allowable Stresses

• S0 - Equal to the higher of S values from Section II-D, Subpart 1, Table 1A, or 300,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> Smt

• Sm - From Section II-D, Table 2A, Sm values at lower temperatures, extended to higher temperatures using same criteria

• Smt is the lower of Sm (time-independent) and St (time-dependent) 13

HBB - Basis for St (HBB-3221)

• The lowest of:

(a) 100% of the average stress required to obtain a total (elastic, plastic, primary, and secondary creep) strain of 1%;

(b) 80% of the minimum stress to cause initiation of tertiary creep; and (c) 67% of the minimum stress to cause rupture (Sr).

• Determination of St is inherently conservative because of the 80% and 67%

factors applied to tertiary creep initiation and stress-to-rupture.

14

Other Stresses/Material Properties

• Sy - yield stress as function of temperature

• Su - ultimate strength

• R - Weld strength reduction factors

• Tensile and yield strength reduction factors for longtime services (Table HBB-3225-2)

• Isochronous stress-strain curves (ISSCs) 15

Deformation Controlled Quantities (HBB-T) 16

• All Class A materials

• Rules found in Nonmandatory Appendix HBB-T

• Bounding analysis Elastic analysis

• All Class A materials

• Rules found in NMA HBB-T

• But no material models in Code (currently)

• Exact analysis Inelastic analysis

• Subset of materials (304 and 316 SS, A617, soon to be Grade 91)

• Rules in N-861 and N-862

• Bounding analysis Elastic perfectly-plastic analysis (EPP)

Characteristics A subset of the design limits:

- Strain accumulation

- 1% average strain

- 2% linearized bending

- 5% maximum strain

- Creep-fatigue

- Buckling Typically are driven by secondary (self limiting) stresses Evaluation Methods

Creep-fatigue (HBB-T-1411)

• Basically:

1. Compute creep damage based on life fraction:

2. Compute fatigue damage based on a cyclic life fraction:

3. Consult interaction diagram for pass/fail

• Welds: same interaction diagram, factors on damage 17 Creep damage Fatigue damage

Creep Damage (HBB-T-1433)

• Construct a stress relaxation curve for each hold in each cycle type

• Determine creep damage with a time fraction rule for each time interval

=1

• Sum creep damage for all time intervals needed to represent the specified elevated temperature service life = =1

(

)

/

• Database: creep rupture tests

• Welds: use stress rupture factor to reduce the creep rupture strength of the base metal time stress Stress relaxation profile Minimum stress-to-rupture for Alloy 617 18

Buckling and Instability (HBB-T-1500)

Limits for both time-independent (creep not significant) and time-dependent (creep-significant) buckling are provided.

Load factors for both load-controlled and strain-controlled bucking provided.

Figures provide temperature/time combinations below which the time-independent buckling limits may be used.

For conditions where stain-controlled and load-controlled buckling may interact, or significant elastic follow-up may occur, the load factors for load-controlled buckling are also to be used for strain-controlled bucking.

19

Elastic, Perfectly Plastic (EPP) Analysis Use different allowable stresses as pseudo yield stress in EPP finite element analysis to determine different bounding characteristics for different failure modes Intended as simplified screening tools in place of elastic analysis methods No stress classification Any geometry or loading Accounts for redundant load paths Simpler to implement

- Based on finite element results at integration points, no linearization Current status EPP Design Check EPP Code Case Materials Currently Covered Primary Load Under development All Class A materials Strain Limits N-861 304H, 316H, Grade 91, Alloy 617 Creep-fatigue N-862 304H, 316H, Grade 91, Alloy 617 Grade 91, Alloy 617 covered by revision of code cases. Not reviewed by NRC 20

Inelastic Analysis Methods Currently the Code does not provide reference inelastic models for any of the Class A materials

• Specification of the material model left to owners Design Specification or designers

• Limits application of the inelastic rules Historical experience on the Clinch River Breeder Reactor Project shows that inelastic analysis is:

• The least over-conservative of the Division 5 options

• Necessary in critical locations where design by elastic analysis is too conservative to produce a reasonable design Current status

• Unified viscoplastic constitutive models for 316H stainless steel and Grade 91 steel have been developed

• Action to add Grade 91 model just balloted.

21

Class B Rules

• Essentially reference III-1, Class 2 rules HCA - Class B Low Temperature

• Allows more materials than HBB

• Mandatory Appendix HCB-II contains allowable stress values

• Different allowable stresses for:

• Negligible creep

• Non-negligible creep

• Mandatory Appendix HCB-III defines times and temperatures where creep effects can be neglected.

HCB - Class B High Temperature 22

Class B Rules Extend rules of Division 1, Class 2 (Subsection NC) to elevated temperature service.

Based on a design-by-rule approach. Design Lifetime concept is not used.

Allowable stresses based on extrapolated 100,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> creep-rupture properties.

Fatigue damage from cyclic service is addressed only for piping with creep effects (HCB-3634).

23

Core Supports

• mainly references Section III-1 rules.

HGA-Low Temperature

• Same materials and allowable stresses.

HGB - Similar to HBB rules.

24

Construction Rules For Nonmetallic Components (Class SN)

• Section III Division 5 is the only design code that provides construction rules for graphite.

• Graphite materials are used in thermal spectrum advanced reactors because of their excellent neutron moderation properties 25

There is no single nuclear grade of graphite -

therefore, cant design around a specific nuclear grade as metals can (i.e., 316H)

Graphite is heterogeneous by nature, and contains significant pores and cracks.

Graphite is not ductile - Brittle or quasi-brittle fracture behavior Graphite Irradiation significantly alters the graphite behavior - Behavior is completely different before and after turnaround dose is achieved.

26

• Because all graphite is brittle and contains preexisting flaws,

• Core components need to be designed to accept some amount of cracking.

Probabilistic versus deterministic design approach Deterministic is generally too limiting for a brittle material A distribution of possible strengths in a material is needed for quasibrittle materials (i.e., flaw size for graphite).

Probability of failure in component based upon inherent strength of graphite grade and applied stresses during operation.

50X 100X 200X 500X ASME Code Considerations 27

1. Simplified Analysis Method

• Simplified Assessment (HHA-3220)

- Simplified conservative method based on ultimate strength derived from Weibull statistics.

• Full Assessment (HHA-3230)

- Weibull statistics for failure probability

- Maximum allowable probability of failure defined for three Structural Reliability Classes (SRCs).

• Design by Test (HHA-3240)

- Full-scale testing to demonstrate that failure probabilities meet criteria of full analysis.

Structural Integrity Assessment Methods 28 Structural Reliability Class Maxi. Prob. of Failure SRC-1 1.00E-04 SRC-2 1.00E-02 SRC-3 1.00E-01 Graphite code is a process.

How to apply degradation to POF From Dr. Mark Mitchell - PBMR Inc.

Degradation Irradiation Designer should determine the specific changes for their selected graphite grade

• Oxidation (HHA-3141)

- Loss of strength and geometry changes to be considered

• Irradiation (HHA-3142)

- Property changes to be addressed

• Abrasion and Erosion (HHA-3143)

- To be considered when there is relative motion or high gas flow rate in gas-cooled designs Special Considerations in Design of Graphite Core Components

Graphite Degradation (Form MDS-1 Material Data Sheet)

ASME BPVC Data sheets capture:

• Material properties

- Strength

- Elastic modulus

- CTE

- Conductivity

- Thermal conductivity (Diffusivity)

• Irradiation effects

• Temperature dependence

- Temperature affects everything

• Oxidation effects

Summary 31 Division 5 was issued as part of the 2011 Addenda to the 2010 Edition of the BPV Code Though the design rules development for metallic components traced all the way to the 1960s Division 5 covers the rules for the design, fabrication, inspection and testing of components for high temperature nuclear reactors Construction rules for both metallic and nonmetallic components are provided The rules for nonmetallic components are unique among all design codes world-wide ASME Code committees are actively pursuing code rules improvement and developing new technologies to support Advanced Nuclear

NRC Review and Potential Endorsement of ASME BPVC,Section III, Division 5 Advisory Committee for Reactor Safeguards July 20, 2021 Jeff Poehler, Sr. Materials Engineer Reactor Engineering Branch Office of Nuclear Regulatory Research Jordan Hoellman, Project Manager Advanced Reactor Policy Branch Office of Nuclear Reactor Regulation

Purpose 33 Provide an overview of the process for NRCs review and potential endorsement of 2017 ASME BPVC Section III, Division 5, High Temperature Materials (Section III-5)

Discuss likely exceptions and limitations to NRCs endorsement.

NRC Guidance Documents for Section III-5 Endorsement 34 NUREG-2245 Technical Review of the 2017 Edition of ASME Section III, Division 5, High Temperature Reactors

• Document the staffs technical evaluation of the 2017 Edition of Section III, Division 5 and Code Cases N-861 and N-862 for acceptability and endorsement. Provide technical basis for DG-1380.

Regulatory Guide (RG) - Acceptability of ASME Section III, Division 5, High Temperature Reactors (DG-1380)

• Describes an approach that is acceptable to the NRC staff to assure the mechanical/structural integrity of components for use in in elevated temperature environments, which are subject to time-dependent material properties and failure modes.

• Contains exceptions and limitations to the staffs endorsement.

• The regulatory guide will update the guidance of RG 1.87.

• Appendix A of DG-1380 contains staff guidance on quality group classification for high-temperature reactors.

Scope of Staff Review Section III-5, 2017 Edition

• Did not review Nonmandatory Appendix HBB-Y, so not endorsing.

1 Code Cases N-861 and N-862 2

Alloy 617 Code Cases

• Separate technical basis document being developed

• Will merge results into final DG-1380 3

35

Contractor Expert Recommendations

• To ensure an independent review of the technical adequacy of Section III, Division 5, NRC used contractors not directly involved with Division 5 code development

• NRC also used contractors more involved with code development on a limited basis to provide historical perspective on Division 5.

36

Review Process -

General 37 Relied on previous reviews when possible.

- Code Cases 1592-1596.

- Section III, Division 1.

The NRC staffs review was augmented by input from several national laboratories and commercial contractors.

See NRCs Advanced Reactor Public Website:

https://www.nrc.gov/reactors/new-reactors/advanced.html#endorev

38 Contractor Topics ML #

PNNL Design, Fabrication, Examination, Testing (HBB/HCB/HGB-3000, 4000, 5000, 6000)

Mechanical design appendixes for metallic core supports (HGB-I, HGB-II, HGB-III, HGB-IV)

ML20269A145 ORNL Materials (HBB/HCB/HGB-2000)

Tables and Figures (Mandatory Appendix HBB-I-14)

Guidelines for Restricted Material Specifications (Non-Mandatory Appendix HBB-U)

ML20269A125 NUMARK

/EMC2 Mechanical Design Appendixes for Class A and Class B components (HBB-II, HBB-T, HCB-I, HCB-II, HCB-III)

ML20349A003 Technical Requirements - Graphite Materials and Design ML20358A145 Code Cases N-861 and N-862 (all aspects)

ML20349A002 ANL Historical Context and Perspective on Materials Properties ML21090A033 Contractor Reports

Review Process - General Requirements Staff compared the 2017 Edition of ASME Code III-5-HAA and -HAB to the 2017 Edition of ASME Code III-NCA to ensure consistency with what the NRC has endorsed in 10 CFR 50.55a.

Exceptions or limitations proposed where there are differences.

39 Similarly, the staff compared the 2017 Edition of ASME Code III-5-HAA and -HAB to the 2019 Edition of ASME Code III-5-HAA and -HAB to ensure consistency with those items that were corrected in the 2019 Edition.

General Requirements

- Examples of Exceptions/Limitations Limitation: Staff does not endorse use of a Certifying Engineer who is not also a Registered Professional Engineer.

Basis: Consistency with a similar condition in 10 CFR 50.55a on 2017 Edition of Section III-NCA.

Limitation: When using HAB-3126(b), HAB-3127(b),

and HAB-3855.3(c)(2) and (d)(2): The procurement documents should specify that the service will be provided in accordance with the accredited ISO/IEC 17025 program and scope of accreditation.

Basis: This is one of several limitations included for consistency with the updated ILAC accreditation process that is called out in NCA-3126 and also in the 2019 edition of Section III-5.

40

Mechanical Design - Exceptions and Limitations

• The staff identified exceptions and limitations related to mechanical design (HBB-3000, HBB-T) for several reasons:

- Consistency with Section III-1 conditions in 10 CFR 50.55a

• Socket weld design condition.

- Consistency with RG 1.87 conditions on Code Case 1592 -

• Use of strain-controlled buckling factors.

- Lack of guidance in Section III-5

• Inelastic analysis for meeting HBB-T deformation limits.

• Stress relaxation cracking.

41

Mechanical Design -

Exceptions and Limitations -

Stress Relaxation Cracking Limitation:

When using HBB-T-1710 applicants and licensees should develop their own plans to address the potential for stress-relaxation cracking in their designs.

Basis:

Stress relaxation cracking is a mechanism causing enhanced creep crack growth in certain materials caused by relaxation of weld residual stresses in components in high-temperature service. Section III-5 does not contain any provisions addressing stress-relaxation cracking.

42

Review Process -

Metallic and Graphitic Materials

• Did not primarily rely on previous reviews.

• Independent analysis of materials properties and allowable stresses by NRC contractor.

• Additional input by subject matter experts familiar with the development of Section III-5.

Class A Metallic materials (HBB-I-14)

• Did not rely on previous reviews.

• Graphite provisions were not in 159X Code Cases - New to Section III-5.

• Technical review of Section III-5 by subject matter experts.

Graphite (HHA) 43

44 Metallic Materials In some cases, contractor independent analysis determined properties and allowable stresses with lower values than the code, suggesting code values are nonconservative.

Lower values were typically only at higher temperatures and longer times for time-dependent properties.

NRC staff considered these findings in a holistic manner, including how these properties are used, inherent conservatism of the Division 5 design rules, and historical context.

Input from ANL provided historical context and perspective on materials properties.

Metallic Materials - Exceptions and Limitations

• For time-dependent allowable stresses, staff placed limitations on endorsement for several materials.

• Limitations in form of maximum temperature limit for several materials.

45 Material Properties Temperature Limit Type 304 Smt, St, Sr 1300 °F, 700 °C Type 316 Sr 1300 °F, 700 °C 2-1/4 Cr-1 Mo Smt, St, Sr 950 °F, 510 °C

• For 9Cr-1Mo-V, 2019 Section III-5 properties are endorsed in lieu of 2017 Section III-5 properties.

Example of Basis for Conditions on Allowable Stresses For Type 304, ORNL independent analysis suggested significant non-conservatism of Section III-5 St values for most times and temperatures. At 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, non-conservatism was suggested at temperatures 850 °F or 450 °C. This is based on independent analysis values more than 10% lower than Section III-5 values.

Most of the apparent non-conservatism driven by the tertiary creep criterion for St.

Tertiary creep criterion for St is a known issue in the Code. It was not intended that this criterion should control most time-dependent allowable stresses.

ANL performed an alternate analysis using a different approach for tertiary creep data. This analysis showed significant non-conservatism only at temperatures 1300 °F or 700 °C.

46

Graphite Materials and Design

• Numark Associates Inc. provided a technical assessment of Subsection HH, Class A Nonmetallic Core Support Structures, Subpart A, Graphite Materials.

• Staff has completed the review of the above report and all applicable sections of ASME Section III, Division 5 and obtained clarifications and feedback from NRC contractors (NUMARK and INL) in order to come up with the conclusions identified in the NUREG.

• The staff's independent review of the code requirements considered the holistic design of graphite core support structures.

47

Graphite Materials and Design -

Exceptions and Limitations Paragraph Parameter Limit in Section III-5 HHA-3141, Oxidation Weight Loss Limit 30%

HHA-3142.4, Graphite Cohesive Life Limit Cohesive Life Limit

+10%

HHA-3143, Abrasion and Erosion Gas Flow Velocity 100 m/s (mean)

HHA-4233.5, Repair of Defects and Flaws Allowed repair depth 2 mm (0.079 inch) 48 Limitations identified by staff where Division 5 has a numerical parameter limit, but staff not convinced the limit is generically applicable to all designs. Design-specific justification is requested for the parameter value in these case:

Graphite Materials and Design - Other Exceptions and Limitations Limitation: The NRC staff is not endorsing the provisions of HHA-3330(g).

Basis: HHA-3330 (g) allow for access to performing inservice inspection. If necessary, inservice inspection may be replaced by operational monitoring Staff is not endorsing this provision because requirements for inservice inspection are outside of the scope of Section III-5, HHA.

The provision related to operational monitoring is the one that the staff finds out of scope.

49

Four Quality Groups and associated standards (from DG-1380, Appendix A)

Quality Group A

• Safety-related SSCs

• Use ASME Section III, Division 5 Class A for safety related SSCs that have safety significance Quality Group B

• Safety-related SSCs

• Use ASME Section III, Division 5 Class B for safety related SSCs with low safety significance Quality Group C

• Non-safety-related SSCs with safety significance

• Use ASME Section VIII, Division 1 or 2 Quality Group D

• Non-safety-related SSCs with no special treatment

• Owner to establish standards for use

Summary Exceptions and limitations were generally identified when the staff found that additional guidance was needed to augment the provisions of Section III-5, or where material properties and allowable stresses are potentially nonconservative.

The NRC staff has completed its initial review of Section III-5 for potential endorsement.

The NRCs review is documented in NUREG-2245.

DG-1380 contains the staffs regulatory position on Section III-5, including some exceptions and limitations.

51

Next Steps The NUREG and DG will be issued for public comment.

Alloy 617 Code Cases technical review (in progress).

Make changes as necessary to NUREG and DG to address public comments.

Reissue DG for a second public comment period incorporating Alloy 617 results and resolution of public comments.

Issue draft Alloy 617 technical basis document concurrently with DG.

52