ML21153A446
| ML21153A446 | |
| Person / Time | |
|---|---|
| Issue date: | 05/05/2021 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Burkhart, L, ACRS | |
| References | |
| NRC-1500 | |
| Download: ML21153A446 (217) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Docket Number: (n/a)
Location: teleconference Date: Wednesday, May 5, 2021 Work Order No.: NRC-1500 Pages 1-149 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 685TH MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 WEDNESDAY 9 MAY 5, 2021 10 + + + + +
11 The Advisory Committee met via 12 Videoconference, at 2:00 p.m. EDT, Matthew W. Sunseri, 13 Chairman, presiding.
14 COMMITTEE MEMBERS:
15 MATTHEW W. SUNSERI, Chairman 16 VICKI BIER, Member 17 DENNIS BLEY, Member 18 CHARLES H. BROWN, JR. Member 19 VESNA B. DIMITRIJEVIC, Member 20 GREG HALNON, Member 21 WALTER L. KIRCHNER, Member 22 JOSE MARCH-LEUBA, Member 23 DAVID A. PETTI, Member 24 JOY L. REMPE, Vice Chairman 25 PETER RICCARDELLA, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 ACRS CONSULTANT:
4 DESIGNATED FEDERAL OFFICIAL:
5 DEREK WIDMAYER 6
7 ALSO PRESENT:
8 CYRIL DRAFFIN, USNIC 9 SCOTT MOORE, Executive Director, ACRS 10 QUYNH NGUYEN, ACRS 11 WILLIAM RECKLEY, NRR 12 JOHN SEGALA, NRR 13 MARTIN STUTZKE, NRR 14 NANETTE VALLIERE, NRR 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 1 P R O C E E D I N G S 2 2:00 P.M.
3 CHAIR SUNSERI: It is 2:00 p.m. Eastern 4 Time. The meeting will how come to order. This is 5 the first day of the 685th meeting of the Advisory 6 Committee on Reactor Safeguards. I'm Matthew Sunseri, 7 chair of the ACRS.
8 I will now call the roll and confirm a 9 quorum and that clear communications exist. Normally, 10 we would start with Ron Ballinger, but he's not 11 available to attend this week and he has an excused 12 absence, so I'll go to Vicki Bier. And I know that 13 Vicki had contacted me and thought her availability 14 might be a little uncertain this afternoon, so sounds 15 like she's not available either. And that is an 16 excused absence also.
17 Dennis Bley.
18 MEMBER BLEY: I'm here.
19 CHAIR SUNSERI: Charles Brown. Charles 20 Brown. Vesna Dimitrijevic.
21 MEMBER DIMITRIJEVIC: Here.
22 CHAIR SUNSERI: Greg Halnon.
23 MEMBER HALNON: I'm here.
24 CHAIR SUNSERI: Walt Kirchner. Walt 25 Kirchner. Jose March-Leuba. Jose? Dave Petti.
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4 1 MEMBER PETTI: Here.
2 CHAIR SUNSERI: Joy Rempe.
3 MEMBER REMPE: Here.
4 CHAIR SUNSERI: Pete Riccardella.
5 MEMBER RICCARDELLA: Here.
6 CHAIR SUNSERI: And myself. So let me 7 check here. One, two, three, four, five, six, seven.
8 We barely have a quorum.
9 Walt, are you on yet?
10 MR. CORRADINI: I thought Walt said he was 11 coming on at 3 p.m.
12 CHAIR SUNSERI: Oh, that's right. Yes, he 13 did talk to me about that. That's another excused 14 absence. So how about Jose? Jose?
15 MR. NGUYEN: I just let him in, so I think 16 he's here.
17 CHAIR SUNSERI: Jose, are you talking 18 about Jose?
19 MR. NGUYEN: Correct. Charlie is also 20 showing up.
21 MEMBER BROWN: I have got to leave my desk 22 here for a minute. I am here.
23 CHAIR SUNSERI: Okay, you're there.
24 MEMBER BROWN: I've got to take care of 25 myself. I'll be back in about three minutes.
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5 1 MEMBER RICCARDELLA: Walt shows as being 2 here, too.
3 MEMBER BROWN: I just logged in just now.
4 CHAIR SUNSERI: All right. Well, we have 5 a starter quorum now. We'll go ahead and get started.
6 Maybe by the time we get through with the 7 introductions everybody will be present.
8 MEMBER BROWN: Okay, I'll answer present 9 when you call me out.
10 CHAIR SUNSERI: That's fine. All right, 11 so let me just divert a little bit here before I 12 continue. I want to take a moment and recognize the 13 fact that former Commissioner Lyons passed away last 14 week and you've likely seen reports on the multitude 15 of technical accomplishments he made. I unfortunately 16 never had the privilege to work with him during his 17 times as a Commissioner and made several drop-ins and 18 visits with him and I found him to be a very brilliant 19 technically, and of very sound character.
20 And I know there's other members on the 21 committee that knew him much better than me and I'll 22 just pause at this moment to see if anyone wants to 23 say anything.
24 MEMBER REMPE: Sure, Matt. This is Joy.
25 And yes, I was fortunate enough to work with Pete when NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 he was at DOE. Again, there are many kind and 2 wonderful things one can say about Pete, but what 3 always amazes me was his low-key manner and civility 4 and had a very kind of way interacting with people 5 which I greatly appreciated.
6 CHAIR SUNSERI: Yes, I totally agree. He 7 was a special person. He's going to be missed by our 8 industry.
9 I think Walt had a close relationship with 10 him as well and wish he was here to say something, but 11 anyway. All right, anybody else? Thank you for that.
12 The ACRS was established by the Atomic 13 Energy Act as governed by the Federal Advisory 14 Committee Act. The ACRS section of the U.S. NRC 15 public website provides information about the history 16 of the ACRS and provides documents such as our charter 17 bylaws, Federal Register notices for meetings, letter 18 reports, and transcripts of all full and subcommittee 19 meetings including all slides presented at the 20 meeting.
21 The committee provides its advice on 22 safety matters to the Commission through its publicly 23 available letter report. The Federal Register notice 24 announcing this meeting was published on April 7th, 25 2021, and provided an agenda and instructions for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 1 interested parties to provide written documents or 2 request opportunities to address the committee. The 3 Designated Federal Officer for this meeting is Mr.
5 During this week, the committee will focus 6 on the following for the remainder of most of the day 7 or the remainder of the day, we're going to take up an 8 interim letter report on 10 CFR Part 53, rulemaking 9 for licensing of advanced reactors. There will be 10 some staff presentations and then we will get into 11 report preparation following that.
12 Tomorrow morning, we will begin with a 13 White Paper on Fusion which is an informational 14 briefing.
15 Regarding Agenda Item 5 for tomorrow, 16 uprated NuScale standard design approval application 17 update, NuScale has requested that this item be 18 removed from the agenda. NuScale plans to provide an 19 update at a date closer to when the standard design 20 approval would be submitted. The purpose of the 21 meeting was to discuss NuScale's 250 megawatt thermal 22 standard design approval regulatory engagement plan 23 and which proposed four phase review process.
24 Interested members of the public may 25 access the regulatory engagement plan in ADAMS and I'm NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 going to give a number here so I'll pause just a 2 second so that you can ready yourself to copy. The 3 ADAMS number is M as in Mike, L as in Lima, 21047 A as 4 in Alpha, 475. That's the engagement plan and the 5 four phase review process proposal is M as in Mike, L 6 as in Lima, 21112 A as in Alpha 183.
7 I apologize for any inconveniences this 8 may have caused by changing this agenda this way.
9 The other topics that we will take up in 10 under the general topic of report preparation and 11 other committee business and that is the NuScale 12 control room staffing plan. This is a letter report 13 that carried over from the last committee meeting.
14 And this will be our number one priority to get this 15 letter report out this meeting. And we also are 16 updating our bylaws as an action item from our 17 retreat. We will work this item in as time permits.
18 As far as the interim letter on Part 53, 19 there's a lot of work that still needs to go into this 20 letter and I talked to Dennis. My goal will be to 21 have a read in of the draft letter today and then try 22 to get agreement on recommendations and conclusions by 23 the end of the week. If we can complete the letter by 24 the end of the week, that would be the stretch goal, 25 but right now I don't know if that's going to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 achievable or not, but we can get at least through the 2 read through and the agreement on the draft 3 recommendations and conclusions that would be a good 4 position to be in.
5 A phone line, a phone bridge line has been 6 opened to all members of the public to listen in on 7 the presentation and committee discussions. We have 8 received no comments and only one request to make oral 9 statements from a member of public regarding today's 10 session. And this is a request from USNIC and that 11 will come during the comment period following the Part 12 53 presentation.
13 There is also an opportunity for public 14 comment and we have set aside time in the agenda for 15 comments of members of the public attending this 16 meeting. Written comments may be forwarded to Mr.
17 Derek Widmayer, the Designated Federal Officer. A 18 transcript of the open portion of the meeting is being 19 kept and it is requested that speakers identify 20 themselves and speak with sufficient clarify and 21 volume so that they may be readily heard.
22 Additionally, participants should mute themselves when 23 not speaking.
24 Now one small change you noticed during 25 the roll call is we have two new members that have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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10 1 been appointed to ACRS since our last full committee 2 meeting. And I want to welcome Vicki Bier and Greg 3 Halnon to the committee here. I'm going to do a 4 little bit of introductions here. I'll put my camera 5 on for this. Vicki is not here, so I'll save this for 6 tomorrow.
7 But Greg Halnon is an independent nuclear 8 industry consultant who has more than 40 years of 9 experience in the nuclear industry. Mr. Halnon has 10 expertise in all aspects of nuclear plant operations, 11 as well as quality standards, security maintenance, 12 and engineering processes. He currently holds a 13 professional engineering license in two states and has 14 held two senior reactor operator licenses during his 15 career.
16 Mr. Halnon earned a Bachelor of Science 17 degree in engineering from the University of Central 18 Florida with emphasis on mechanical and thermal 19 hydraulics. And it is also worthy to note that Greg 20 is a life member of the American Nuclear Society.
21 So Greg, welcome to the committee. If you 22 have anything you want to say before we get started?
23 MEMBER HALNON: Thank you, Matt. Real 24 briefly. I just appreciate everybody's welcoming and 25 it didn't take me long to very much appreciate the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 1 quality of people, both on the staff and ACRS. And I 2 really look forward to interacting through this 3 appointment. So I appreciate the time.
4 CHAIR SUNSERI: And I'm not doing 5 something right here with my attendee list, but I see 6 somebody has their hand up and I don't know who that 7 is, so whoever has their hand up, you have the floor.
8 MEMBER MARCH-LEUBA: It might be me, 9 sorry. This is Jose. I'm back. I've been taking 10 every single thing so yes, I am back.
11 CHAIR SUNSERI: Could you hear us when we 12 were doing the roll call?
13 MEMBER MARCH-LEUBA: No, I couldn't. It 14 was a long story, but I'm back.
15 CHAIR SUNSERI: Okay, all right. No 16 problem. All right, all right, well, that's good. We 17 have a strong quorum now.
18 All right, so that is all for the 19 introductions and opening remarks. I'll open the 20 floor to the committee. Any member have anything you 21 want to say before we get into the agenda?
22 All right, well, then at this point I will 23 turn the floor over to Dennis Bley for the interim 24 letter report on 10 CFR 53.
25 MEMBER BLEY: Thank you, Mr. Chairman.
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12 1 This is not dealing directly to the letter; the staff 2 at the back to give us a presentation. I have asked 3 them to do another overview especially on Subparts B 4 and C for members who were not able to listen in all 5 of our subcommittee meetings.
6 I've asked the staff to do this as quickly 7 as reasonable to do that review. There were also some 8 areas that came up at our April 22nd meeting where 9 some additional staff's expertise would have been in 10 response to questions from the committee and I've 11 asked them to go back over some of those issues so 12 that they're planning to do that.
13 After we get through the staff 14 presentation, you may know that USNIC has asked for a 15 chance to speak and Mr. Cyril Draffin will then 16 provide comments to us on their behalf. And then 17 we'll do a read through of the letter that's been put 18 together to support this thing and we'll try to make 19 sure we get through all of that before the end of the 20 day.
21 So at this time, I'm going to turn it over 22 to staff.
23 John Segala, did you want to begin or 24 somebody else?
25 MR. SEGALA: Yes, thank you. Yes, this is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 1 John Segala, Chief of the Advanced Reactor Policy 2 Branch in the Office of Nuclear Reactor Regulation.
3 And consistent with the Nuclear Energy Innovation and 4 Modernization Act, or NEMA, we are committed to 5 developing a technology inclusive, risk informed, and 6 performance based regulatory framework for a wide 7 range of advanced reactor designs and publishing the 8 final Part 53 rule by October of 2024 in accordance 9 with the Commission's directed schedule.
10 We are committed to a regulatory framework 11 for advanced reactors that achieves the goals of the 12 Commission's advanced reactor policy statement and the 13 NRC's principles of good regulation. We are having 14 extensive stakeholder engagement where we release 15 preliminary rule language to solicit feedback to 16 better inform the staff's proposals and to ensure a 17 shared understanding of what will be included in the 18 final rule.
19 As we are considering changes to the 20 previously released preliminary rule language, we want 21 to ensure that we have appropriately considered the 22 feedback we have received from all stakeholders 23 including the public, industry, standards development 24 organizations, trade groups, non-governmental 25 organizations, and the Advisory Committee on Reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 1 Safeguards.
2 Since we are at the early stages of the 3 rulemaking process, the draft preliminary rule 4 language will remain open for discussion as the staff 5 works towards providing the Commission a proposed 6 rule. We are here today in the fifth of many ACRS 7 meetings we will be having this year to seek ACRS 8 feedback on the NRC's development of Part 53 9 preliminary proposed rule language for advanced 10 reactors.
11 We previously briefed the ACRS 12 Subcommittee in January on the first set of 13 preliminary rule language in Subparts B and F, in 14 February on Subparts C and D in March, where 15 stakeholders shared their insights and we discussed 16 the structure and logic of Part 53, key guidance 17 needed for Part 53 and Subpart E on construction and 18 manufacturing.
19 In April, our last meeting where we 20 discussed the second iteration of the preliminary rule 21 language in Subparts B and C and the key elements of 22 the Part 53 framework in order to set the stage for 23 the ACRS full committee meeting today.
24 Today, we plan to provide the full 25 committee an overview of the Part 53 structure and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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15 1 preliminary rule language for Subpart B and C. We 2 also plan to provide additional information to help 3 answer questions brought up during the April 4 subcommittee meeting. We understand that ACRS plans 5 to develop an interim letter report following this 6 full committee meeting and we are looking forward to 7 hearing any insights and feedback from the full 8 committee today, as well as the conclusions and 9 recommendations in the ACRS interim letter.
10 This completes my opening remarks. And I 11 can turn it over to Bill Reckley or Bob Beall.
12 MR. RECKLEY: Yes, this is Bill Reckley.
13 Did you have something, Dennis?
14 MEMBER BLEY: No, I'm conferring, but I 15 would like to hear from you though.
16 MR. RECKLEY: All right. We can go to the 17 next slide.
18 So on Slide 2, as John mentioned, we're 19 going to go over the overall structure. We're not 20 going to spend too much time on that. There seem to 21 be an general understanding and at least for now a 22 general support of the overall structure. Then we're 23 going to look at Subpart B on the safety requirements 24 and Subpart C design and analysis.
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16 1 of material to try to go through in a couple of hours.
2 So we're going to go relatively quickly. If there's 3 a need to stop and pause and go over some things, that 4 would be understandable. But some of it also, some of 5 the specific topics like the probabilistic risk 6 assessment, some elaboration on our plans to use PRA 7 within Part 53, we have added a few slides and Marty 8 Stutzke will be doing that presentation when we get 9 into Subpart C on design and analysis.
10 But for now, if we go to slide -- the next 11 slide. John mentioned this in the background already.
12 We had, as part of overall plans for advanced 13 reactors, considered a rulemaking even back in 2016, 14 as we were laying out our strategies. We were -- then 15 events kind of overtook us with the passage of NEMA 16 and signing that into law in 2019 and that law 17 specifically told us to develop a framework through 18 rulemaking to address advanced reactors and so that 19 changed our schedule a bit and is largely the reason 20 we're here today.
21 If we go to Slide 4 --
22 MEMBER REMPE: Bill?
23 MR. RECKLEY: Yes.
24 MEMBER REMPE: This is Joy. I thought of 25 something when we last met that Dennis and I discussed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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17 1 later. And I think Dennis said he didn't hear what I 2 heard and I'm not, maybe I got confused. But I asked 3 you a couple of questions last meeting about mostly 4 gatekeeper, you were trying to do with Part 53 giving 5 folks a little more flexibility because they are an 6 advanced reactor with increased reliance on passive 7 and inherent safety features.
8 So what if someone comes with something 9 that's a Superphenix, you know, Clinch River thing 10 that doesn't have any passive, well, not many, passive 11 or inherent safety features. What's the gatekeeper --
12 because I thought you'd said well, if we see something 13 like that and it's not meeting the requirements, we're 14 going to impose some additional requirements on them.
15 It sounds kind of fuzzy to me.
16 When do you decide it's got sufficiently 17 increased reliance on passive and inherent features?
18 That sounds a little fuzzy.
19 MR. RECKLEY: And to be clear, NEMA gave 20 a number of criteria and some of it was related to 21 passive or inherent safety features. Other of the 22 criteria within NEMA had to do with cost of 23 electricity, fuel utilization, or waste yields, 24 reliability, proliferation, increased thermal 25 efficiency which most or at least many of the non-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 1 lights would engender, or the ability to use for non-2 electric applications like hydrogen production. So 3 within the NRC's advanced reactor policy statement, 4 the focus was, as you mentioned, passive inherent 5 safety benefits.
6 Within NEMA, there's a number of 7 considerations that could qualify one to be quote an 8 advanced reactor. So we had in our rulemaking plan 9 acknowledged that light water at a minimum, light 10 water SMR, Small Modular Reactors, and any non-light 11 water reactors, so a generation for technology be it 12 a Superphenix or a medium size fast reactor like PRISM 13 or some of the micro reactor designs being considered 14 now. Any of those would have been falling into that 15 category most likely.
16 The question had become what about large 17 light water reactors, the Generation III+ kind of 18 technologies. Our original thinking was they may or 19 may not come into play. So at this point, you know, 20 there will be many more that would go through the gate 21 than would be stopped by the gate. Maybe I can just 22 put it that way, if that answers the question?
23 MEMBER REMPE: What's the gate? I recall 24 a long time ago that the Commission said well, the 25 current fleet is safe enough and advanced reactors NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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19 1 don't have to be safer because the current fleet is 2 safe enough. And we didn't give them increased 3 flexibility, so it would sound to me is what you're 4 saying is everything gets in or is there some place 5 where if they meet the safety criteria everybody is 6 in, right?
7 MR. RECKLEY: Yes, pretty much. The only 8 question that really NEMA raised when you look at 9 those criterias and this is just from my point of 10 view, the only question raised is it said advanced 11 reactors other than those under construction at the 12 time of the act. And the only construction under way 13 at the time of the act was AP1000.
14 So that's why we were questioning whether 15 Gen III+ might be included, but all of this will be 16 kind of brought out as we finish out this rulemaking 17 and agree on the scope. But there won't be very many 18 technologies excluded in my mind based on the criteria 19 that Congress included which went well beyond passive 20 or inherent safety features and included things like 21 fuel utilization, non-electric uses and so forth.
22 MEMBER REMPE: Then I guess my next 23 question is why do you keep saying these things are 24 going to be safer because you just had told me 25 everybody is in?
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20 1 MR. RECKLEY: In the advanced reactor 2 policy statement, first of all, I hope we don't 3 overuse that they are safer, as you mentioned, and as 4 we'll get into as we go through Subparts B and C. The 5 way I like to put it is they provide their safety 6 through different mechanisms and light water reactors 7 include a certain amount of reliance, for example, on 8 mitigation, including emergency planning, siting 9 restrictions, and so forth.
10 One of the goals of Generation IV reactors 11 and we've heard from stakeholders that this remains 12 true is to lessen the reliance on things like siting 13 and emergency planning as a safety measure and as it 14 was discussed in the advanced reactor policy 15 statement, to ensure safety more through the design of 16 the facility and the use as you've mentioned a couple 17 of times, the use of passive and inherent safety 18 features.
19 So when you look at it as an integrated 20 assessment of the safety of any particular plant down 21 the road, the overall safety in terms of protecting 22 public health and safety will be at least as good as 23 what we have now for the light water reactors, but 24 more importantly how you get that safety might be 25 different.
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21 1 MEMBER REMPE: It might be different, but 2 it doesn't have to be different.
3 MR. RECKLEY: It doesn't have to be 4 different.
5 MEMBER REMPE: Yes. I think we need to 6 keep this in mind because making a lot of assumptions 7 about oh, they're going to do more things with passive 8 and inherent, but unless there's a gate to keep that 9 and what is more reliance, there isn't any.
10 MR. RECKLEY: Right.
11 MEMBER REMPE: So that's like of a fiction 12 here, it's all everybody's dream, but anybody can get 13 through the gate is I guess what I'm learning a little 14 more explicitly.
15 MR. RECKLEY: And another way to put that 16 is, for example, we're allowing, we're trying the way 17 we're writing the rule, to allow for the use of less 18 reliance on emergency planning if it can be justified.
19 But we're not requiring that there be no reliance on 20 emergency planning, right?
21 So going to exactly what you said, they 22 could achieve safety through some of the same 23 measures, siting restrictions, and emergency planning, 24 and so forth as the current fleet. And we're trying 25 not to preclude it, but we're also saying if in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 1 design process you can justify that you don't need to 2 rely on those measures, then we're trying to build in 3 the flexibility to say that you've proved your point 4 through the design process and you don't need those 5 additional measures that were imposed for light water 6 reactors.
7 MEMBER REMPE: Thank you. I appreciate 8 this long discussion on this, but I think it's good to 9 lay it out on the table.
10 MR. RECKLEY: And it's actually, thank 11 you, because it's kind of important as we go to the 12 next -- we can go to the next slide on Slide 4.
13 MEMBER BROWN: Bill, Bill?
14 MR. RECKLEY: Yes.
15 MEMBER BROWN: I just want to echo Joy --
16 go backwards again.
17 MR. RECKLEY: Okay.
18 MEMBER BROWN: You make the comment, this 19 is what bothered me the whole time on these advanced 20 reactors, they're safer, they're passive. We've heard 21 that continually and they're uranium. It's got to be 22 fissioned in order to produce power. And you won't 23 produce power for large populations, a lot of power in 24 many cases, most cases. So you've got all the same --
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23 1 differently.
2 How in the world can you ever get away 3 from emergency planning zones and site boundaries and 4 all that other kind of stuff and/or dosage 5 requirements? I don't see how you could ever -- it's 6 all the same stuff. All we're doing is adding more 7 toxic means in most circumstances, lead bismuth, 8 sodium. I'd love to have one of those plants go melt 9 down somewhere.
10 So the idea that they're passive and that 11 makes them safer just means it's less complex to be --
12 to make sure the plant shuts down. So I hate the 13 advertising of we're going to get rid of emergency 14 planning zones and everything else. It's just an 15 observation. I had to see it used. So I just needed 16 to say my piece as well.
17 I'm not criticizing you, Bill.
18 MR. RECKLEY: No, that's fine. And again, 19 we're just trying, and we'll get into this a little 20 later in the discussion, trying to take this 21 integrated look and say when would you rely on certain 22 provisions and when might you justify that you don't 23 need to rely on those provisions. We'll get into that 24 a little later as we go along.
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24 1 discussion of how safety is achieved is exactly what 2 the Gen IV International Forum and the experts 3 internationally thought about as they rolled out the 4 leading Gen IV concepts. So I see a lot of 5 consistency with that, thank you. Thanks.
6 MR. RECKLEY: Okay, thanks, Dave. Okay, 7 so if we can go to Slide 4. This is our overall 8 structure for Part 53 and how we arranged it into a 9 number of subparts.
10 And most of the discussion with the 11 subcommittee thus far has been on Subparts B and C.
12 But the general arrangement is that Subpart B was 13 intended to layout the safety goals, the safety 14 objectives, the criteria that would be used and the 15 need to identify safety functions.
16 And then the other subparts were basically 17 organized along the lines of a project lifecycle and 18 were intended to, for example, under design and 19 analysis, Subpart C say, what is the contribution that 20 design and analysis provides to make sure that any 21 particular plant, any particular design meets the 22 safety objectives and meets the safety criteria.
23 And then likewise, what's the role of 24 citing, construction, operations and how would that be 25 carried into retirement. So that was the nature of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 1 the main technical subparts, B through G.
2 Then in addition, there were subparts 3 related to licensing that we're currently developing.
4 Those are Subparts H and I. And then administrative 5 reporting, other general provisions, that would be in 6 Subparts A and J.
7 But again, most of the focus with the 8 Subcommittee, and with external stakeholders, has been 9 on Subparts B and C, up to this point.
10 So, if we got to Slide 5, another way to 11 lay this out for the Subcommittee. We had gone 12 through the individual chapters. For the sake of time 13 I didn't do that today, but this just lays out the 14 same thing from the graphic in kind of more of a chart 15 or a table of contents with the Subparts A through J.
16 Subparts B and C are highlighted because 17 we want to spend more time talking about those today.
18 In red is just some of the, some notes on particular 19 subparts.
20 And in particular, on Subpart B, under 21 safety criteria, some of the discussion topics as 22 included. Our organization of the requirements into 23 the first and second tiers, or categories, of safety 24 objectives.
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26 1 within the criteria related to normal operations. And 2 how you achieve defense-in-depth.
3 And then under the design and analysis, 4 some discussion on how we propose to have the design 5 criteria addressed and the role of probabilistic risk 6 assessments that we'll get into in a few minutes.
7 So, I had not planned on spending much 8 more time. If we can just go back to Slide 4 for a 9 second.
10 I hadn't planned to spend much more time 11 on the overall structure or organization of Part 53, 12 unless there are specific questions.
13 MEMBER BLEY: Yes. This is Dennis, Bill.
14 Charlie brought up a point and argued it, and it's 15 similar to, Charlie, it's similar to what you wrote 16 down for, and delivered it in the last meeting last 17 week.
18 And essentially it boils down to, you 19 cannot assume that new reactors coming in will be 20 safer. And the objective to that assumption.
21 And it seems to me, although some of the 22 hype makes that assumption the approaching you're 23 taking doesn't allow for it. You can't assume it, you 24 have to show that you meet the level of safety. You 25 can comment on that or not?
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27 1 MR. RECKLEY: Well, I think as we're 2 trying to, as we get into Subpart B, on the criteria, 3 the other thing that we will emphasize is that based 4 on past Commission decisions, the highest level 5 criteria remain the same. We haven't proposed, for 6 example, and we'll get into the discussion on the 7 health objectives, but we haven't proposed to use 8 different health objectives, we're using the same ones 9 from the advance reactor policy statement.
10 Again, how you achieve those objectives 11 might differ from design to design. In terms of the 12 plant design, there is going to be reliance on 13 different barriers and technologies based on the type 14 of reactor.
15 And again, we're laying out the 16 possibility that if any designer or licensee wanted to 17 use mitigation measures, the comparable up to what 18 light water reactors do, then we're not precluding 19 that. So, I'm not sure I addressed your question, but 20 --
21 MEMBER BLEY: That's fine, Bill.
22 MR. RECKLEY: Okay. All right, so if 23 there is no other discussion of the overall structure, 24 we can get into Subpart B. Yes, thank you.
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28 1 that I'm not sure we emphasized in the Subcommittee 2 meeting that I did want to just revisit and emphasize 3 by having a slide is that we have said in various 4 papers, including the rulemaking plan sent up to the 5 Commission, SECY-32, that we were planning to build 6 Part 53 based on the activities that were ongoing at 7 that time or that we had completed shortly before 8 then. Such as Secy-19-0117. Which I won't read that 9 long title, but the shorthand of that is licensing 10 modernization project.
11 And NEI 18-04, the ACRS looked at that 12 paper and at the associated reg guide, Reg Guide 13 1.233. But I just wanted to reiterate that our plan 14 was to take such a risk-informed approach. And that's 15 what was communicated to the Commission in those 16 papers and what was accepted within the SRM for both 17 the rulemaking plan and SECY-19-0117.
18 So, if we go to Slide 8. This goes 19 largely to both Charlie and Joy were mentioning. The 20 nature of a reactor is that it's making fission 21 process. It's how it makes its energy, and as a side 22 product it's making fission products. And that's the 23 hazard.
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29 1 safety is basically to provide barriers to the 2 dispersion of those radionuclides to the environment.
3 And in some cases, as we talked about a 4 couple of times, if you cannot preclude the dispersion 5 then you might have mitigation measures on the 6 outside. Such as restricting where you can cite them 7 and/or providing protective actions, such as the 8 sheltering or evacuation in nearby populations. And 9 so, all of those things considered are what determines 10 the risk to public health and safety.
11 We first used this graphic, or a graphic 12 that was similar to it, in SECY-19-0117 to try to 13 describe how within the risk-informed approach we were 14 reflecting in that paper, considers things like 15 mechanistic source term and a more integrated 16 approach.
17 And so, you will see the other paper cited 18 there is the functional containment paper, SECY 19 096. Where if, to simplify a little bit, for light 20 water reactors the general approach has been generally 21 to pick bounding, challenging kind of events for each 22 barrier.
23 And so there would be challenges to the 24 cladding and then there would be challenges to the 25 reactor coolant system and challenges to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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30 1 containment. They may be different events. And the 2 most challenging event basically establishes the 3 design requirements on each barrier.
4 Under this you're still looking, you 5 largely are going to have the same or types of 6 barriers, but you're taking a more, you're looking at 7 more event scenarios and taking an integrated approach 8 to looking at each scenario. And that is reflected, 9 again, the functional containment paper, in the 10 licensing modernization paper, as another way to 11 basically look at ensuring that appropriate barriers 12 are in place to the release of radionuclides.
13 And we'll get into this as we go into the 14 discussion a little more on licensing bases events and 15 Marty's discussion on the use of the probabilistic 16 risk assessments.
17 But the, so, let's go on to Slide 9. This 18 is a slide we used during the Subcommittee meeting.
19 And we look at Part 53 in our construct, 20 and that overall structure, one of the things to keep 21 in mind, just to keep the terminology straight as we 22 go through Subparts B and C, is this --
23 THE OPERATOR: If you're on the bridge 24 line please mute your phone. Please mute your phone.
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31 1 of hierarchy is shown in the chevrons that we start in 2 Subpart B laying out the safety criteria, then we 3 require safety functions as a means to satisfy those 4 criteria.
5 And then when we get into Subpart C we'll 6 talk about design features. Which is the hardware, 7 the structure systems and components needed to carry 8 out the safety function. And then functional design 9 criteria, which are the more specific things 10 associated with the design feature to make sure that 11 it will support meeting the safety functions and the 12 safety criteria.
13 So over in the white boxes, the functions 14 are things like what barrier is needed, what cooling 15 might be needed to maintain a barrier.
16 The design feature would be specific 17 structures and systems and components, pumps, heat 18 exchangers, control rods, whatever the function and 19 then whatever the design feature you're using.
20 And then the design criteria would be 21 things like leak rate, reactivity insertion rates, 22 cooling capacities, more specific engineering 23 parameters associated with components. Like pumps and 24 heat exchangers and so forth.
25 So this is the layout of Subpart B, with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 1 the first, or major focus being the safety criteria 2 and the safety functions. And then in addition we 3 have specific requirements on assessing the unplanned 4 or licensing basis events, ensuring defense-in-depth 5 and the protection of workers.
6 So, another way to characterize Subpart B 7 is these are the, Subpart B is the what. What are we 8 trying to accomplish. And that's, again, meeting the 9 safety criteria, supplying the safety functions.
10 Subpart C, and all the other subparts, get 11 into the how. What are the design features. In the 12 parentheses we start to address things that we will 13 put in requirements in operations related to human 14 actions. What are the role of personnel and so forth.
15 So, the other subparts talk about the how.
16 So, if we go down one more. I repeated 17 this graphic, again, in Slide 10, and enhanced it just 18 a little bit to bring in an example of how, in the 19 past, we've used such an exercise of going from 20 functions to features to functional design criteria in 21 a specific reactor type. And this is the MHTGR.
22 Modular high temperature gas reactor.
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33 1 design criteria that are included in either Appendix 2 A to Part 50 or in Reg Guide 1.232.
3 But you can see that the safety function, 4 as we've currently defined it in Subpart B, and we'll 5 get to the language in a minute, starts off with the 6 ultimate goal of limiting the release or 7 radionuclides. And then identifies what other 8 functions are needed to carry that out.
9 And that is, those functions that are 10 needed to protect whatever barriers a designer is 11 choosing to accomplish those functions. By and large, 12 that's going to be the fuel, some reactor system.
13 Whether it be the fuel encased in the cladding or some 14 kind of pressure boundary. And then in some cases, an 15 additional structure as a last barrier.
16 But for MHTGR it was identified, and we 17 gave the example during the Subcommittee meeting, that 18 you had heat generation or reactivity, heat removal.
19 This is decay heat removal systems, emergency systems.
20 And then for that time frame, when MHTGR was being 21 considered, they identified chemical interactions as 22 a different function.
23 But within our system, if those are the 24 required safety functions, then Part 53 would then 25 also say, within Subpart C, that we'll get to, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 have to identify the design features that are going to 2 accomplish those functions and then you identify 3 additional engineering parameters to show that they 4 can support that.
5 And if you go back and look at the MHTGR, 6 and then to some degree, how that was carried through 7 next generation nuclear plant, you can see how this 8 has fed into the approach that we're proposing for 9 Part 53.
10 So, as the ACRS mentioned in your letter 11 on SECY-19-0117, that methodology and the methodology 12 that you're seeing in Part 53, is really an evolution 13 over the last 30 years. And so, some of what we're 14 going to do today is a little bit of history to kind 15 of fill in where we're getting this.
16 So, if you go to Slide 11, this is 17 basically the same slide again. The top of this slide 18 is right out of MHTGR and NGNP documents that show how 19 you go down and determine those, what are your 20 required safety functions.
21 And then I just added on to that figure 22 for Part 53 space. In Subpart C you would do the 23 design features and the functional design criteria in 24 order to fill out the detail on how you were doing 25 something like removing decay heat.
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35 1 So, if we go to Slide 11. I mean 12.
2 Thank you. One of the things that we talked about, 3 and think is true, is that if you look at the general 4 design criteria and the exercise that was done in the 5 late '60s to develop the general design criteria for 6 light water reactors, and then even more recently 7 three or four, well, from three years ago with the 8 issuance of Reg Guide 1.232 on the advance reactor 9 design criteria, the same exercise that we just 10 described in going from safety functions to design 11 features to functional design criteria, was largely 12 what was done for the, to develop the specific 13 requirements for light water reactors that's reflected 14 in the GDC.
15 And so, if you look at the safety 16 functions in the left in the first column, you'll see 17 reactivity control, fluid systems for heat removal and 18 containment systems. Those align pretty closely to 19 the fundamental safety functions of reactivity heat 20 removal and containment.
21 Or if you're familiar with another one, 22 the three C's, control, cool and contain. So those 23 principles were laid out in the GDC.
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36 1 the individual GDC. So, we think it's a exercise that 2 is similar.
3 In order to support Part 53, which is 4 intended to be a technology inclusive approach, what 5 we're building into Part 53 is a requirement to go 6 through this methodology. And every designer or 7 applicant ultimately would have to do this in order to 8 come up with how they're going to perform the 9 functions and what design features they're going to 10 rely on to do that.
11 So, it is, to some degree, what we're 12 proposing to do on Part 53, and when we get to the 13 actual language in Subpart B, is to replace a fairly 14 perspective list of technical requirements with a 15 methodology to accomplish the same thing.
16 And when we were talking to the Committee 17 during the review of the reg guide and the SECY paper 18 on the licensing modernization project, we had the 19 same discussion of looking at these as a methodology 20 and a requirement to go through this exercise, versus 21 having a prescriptive list because the regulator, or 22 someone else, had already done it.
23 So, a useful exercise, if you have time, 24 is to really look, for example, at the reg guide on 25 the advance reactor design criteria, Reg Guide 1.232, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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37 1 and look specifically at the MHTGR. One of the 2 technologies that's addressed in the advance reactor 3 design criteria is the MHTGR.
4 I point out that it's particularly 5 interesting because the MHTGR was really, from a 6 design and licensing process, the genesis of much of 7 what we're talking about, in terms of licensing 8 modernization, and even moving forward into Part 53.
9 And so, those folks that were involved in that, at the 10 time of the NGNP, were looking at the ARDC and 11 translating and doing this exercise.
12 And so you'll see, through the MHTGR ARDC, 13 some degree of how this plays out. And it somewhat 14 proves the point, at least to me, that the methodology 15 can get you to basically the same place.
16 If we now can go to 13. It just finishes 17 out the rest of the GDC, in terms of the other safety 18 functions. Fluid systems or cooling and containment.
19 And then, we can now start to get into, 20 okay, we have one more slide and then we'll get into 21 the Subpart B actual language.
22 So the next slide, Slide 14, again, just 23 tries to do some comparison of what people are more 24 familiar with, which is the light water or Part 50 25 construct. And some of the changes or alternatives NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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38 1 that are being looked at in Part 53.
2 So, we talked earlier, the safety criteria 3 are basically the same. We're using the same 4 reference values. The 25 rem at the exclusionary 5 boundary. We're using the same QHOs.
6 Albeit, the QHOs don't show up 7 specifically in Part 50, but as we've talked about, 8 they are used in Part 52. Specifically under Chapter 9 19 of the SRP. I think Marty will talk about, more 10 about that when he goes through some of the PRA 11 discussions.
12 Within the design and analysis area, the 13 design basis events are similar. But under Part 50, 14 given the way that Part 50 was developed, it's more 15 prescriptive. It's more conservative.
16 It includes, for example, and we're going 17 to have specific slides on the single failure 18 criterion, so I have it highlighted, but I didn't want 19 to spend much time on this slide.
20 Under Part 53 it's more, it still includes 21 a deterministic DBA in terms of a test of safety 22 related equipment. It still performs that function of 23 having a deterministic DBA.
24 It's probably a little more, it's a little 25 less conservative under Part 53. And the reason for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 1 that is the next bullet. Is that, under Part 50, 2 beyond design basis events were kind of ad hoc under 3 Part 50.
4 And under Part 53 you have a whole 5 category, a new category of events, in which you're 6 doing a methodical assessment. And coming up with 7 design and programs and operator actions needed to 8 address the events down in that category. We will 9 also talk about that when Marty does the PRA 10 discussion.
11 The special treatment for non-safety 12 related, but safety significant SSCs.
13 MEMBER BLEY: Bill?
14 MR. RECKLEY: Yes, Dennis.
15 MEMBER BLEY: I want to stop you just a 16 second. I agree with what you said up there but if 17 one looks at your 53.450, Paragraph F, analysis of 18 design basis accidents, it uses the words conservative 19 and the other typical words. But it doesn't really 20 define what they mean.
21 See, what they mean right now is defined 22 from the SRP in Chapter 15, and you don't have any 23 definition here. Which means some people are kind of 24 feeling empty that there is no requirement.
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40 1 great point to clarify or enroll language or guidance 2 provide.
3 The reason I say it's somewhat less 4 conservative is that the event that's being analyzed 5 under Part 50, and in particular, for example, 6 treating the double ended guillotine break. The 7 frequency of that particular event, if you looked at 8 it from a PRA standpoint, might move it under Part 53 9 down to a beyond design basis event.
10 And it would still need to be addressed 11 but it might now show up as the design basis accident 12 as it was, for good reason, for the light water 13 reactors when it was developed using the process 14 developed for Part 50. The evolution of Part 50.
15 So, in terms of like the thermal 16 hydraulics, and some of the conservatisms that are 17 built in to making sure that if you're using a 18 particular correlation or something like that, then 19 that would be, that's, when we say a conservative 20 analysis under, in our Part 53, Section 450, that's 21 what we were referring to when we say conservative.
22 You know, you'd have to make sure that the 23 actual modeling including the appropriate 24 conservatisms in the DBA. Some of this we'll get into 25 as we discussed the specifics, I think.
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41 1 But again, the reason I highlighted single 2 failure, that came up during the Subcommittee meeting, 3 but we have a couple of slides specifically on that.
4 And also looking at the combinations of failures 5 within the PRA, and the two topics are related, as 6 we'll see as we get into the discussion.
7 So, Slide 15. So now we're going to get 8 into the specific language. On Slide 15, this is the 9 language that we established, that the objectives are 10 to limit the possibility of an immediate threat to 11 public health and safety, and then appropriate 12 measures considering risks.
13 There was a discussion during the 14 Subcommittee meeting on whether we would need to 15 define those terms more.
16 And I just wanted to point out, and have 17 highlighted here, that from our perspective, though 18 the meaning of those terms, if you just read them as 19 they're written, might lead to questions. But the 20 last sentence there is meant to clarify what we mean 21 by that. And these safety objectives shall be carried 22 out by meeting the safety criteria identified in the 23 subpart.
24 So to translate that, what do we mean by 25 an immediate threat to public health and safety, is an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 1 event that would lead to 25 rem over two hours at the 2 exclusionary boundary or over the duration of the 3 event at the low population zone boundary. So that's 4 what we're equating to be an immediate threat to 5 public health and safety.
6 And then as we get to the second tier, 7 what do we mean by appropriate, considering potential 8 risk to public health, that's the QHOs. So, we do 9 think that sentence, hopefully, puts in context what 10 we mean by the high level objectives.
11 If we go to 16, we go into start talking 12 about the first tier. And the language, as I just 13 said, was that the dose, largely from a, well, it's 14 two parts.
15 Part A is normal operations. And we 16 include, within the first tier safety criteria, the 17 100 millirem from Part 20. That's the annual dose 18 from normal effluence.
19 And then more focus is on the unplanned 20 events. And again, we use the same reference values.
21 And as we'll talk about under, as we go down, this 22 analysis is a DBA type analysis, only relying on 23 safety related equipment. And it will show that the 24 dose at the EAB, or low population zone boundary, is 25 less than 25 rem over the duration of two hours or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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43 1 over the whole course of the event.
2 So if we go then to Slide 17, this is 3 somewhat repetitious of what I just said. Twenty-five 4 rems, same reference values we've historically, only 5 relied on safety related equipment demonstrated by a 6 deterministic type DBA.
7 It also is the vehicle for which we ensure 8 an appropriate protection against external hazards.
9 Again, that's largely consistent with how it's done 10 now where the safety related equipment is protected 11 against design basis seismic events or floods or other 12 hazards.
13 And then one last point is, we're going to 14 carry this through, as I mentioned, through the whole 15 rest of the subparts. And it shows up again, for 16 example, under what is the equipment that would be 17 handled and controlled, tightly, through technical 18 specifications. It would be the equipment needed to 19 satisfy this first tier.
20 So the desire there would be to be always 21 able to say that the plant is meeting that first level 22 goal of not presenting an immediate threat to public 23 health and safety.
24 And then as we get into the second tier, 25 you'll see a parallel where we try to take a risk-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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44 1 informed approach. And also carry it through the 2 design, the construction and operations where we 3 basically have the same table for non-safety related 4 but safety significant equipment, and the need to 5 define special treatment for all of those things 6 throughout the lifetime.
7 So, if we go onto Slide 18.
8 MEMBER HALNON: Excuse me, Bill?
9 MR. RECKLEY: Yes.
10 MEMBER HALNON: Bill, this is Greg Halnon 11 and I just wanted to comment on the immediate aspect 12 of this. And we don't have to have a lengthy 13 discussion, but by putting the term immediate in the 14 rule itself, it gives it a very temple emphasis as 15 opposed to the way you're describing it, at least in 16 my mind, is more emphasis on consequence opposed to 17 the tempo aspect of it. So keep that in mind.
18 And when I read it I see a tempo, urgent 19 tempo aspect to it. And the way you described it, at 20 least in my mind is, more of a consequence or an 21 ultimate consequence of an event that could be very 22 long as opposed to intermediate thing. So, anyway, 23 that's my opinion there.
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45 1 got the word immediate threat to public health and 2 safety is actually from the case law on technical 3 specification and what is the appropriate content to 4 tech specs.
5 And knowing that we wanted to carry that 6 threat all the way through, and just looking at how 7 things had been characterized, tech specs, and again, 8 since it's been what we've regulated, large light 9 water reactors, it was put in the terms of an 10 immediate threat to public health and safety.
11 And so, maybe we can look at that. We'll 12 look at that language as we go through the future 13 iterations. I understand what you're saying though.
14 So, if we go then to Slide 18 it lays out 15 the second tier of criteria. And it's been much 16 discussion, but for normal operations, normal 17 effluence, we have kept that they should be kept as 18 low as reasonably achievable.
19 We're looking at future wording to tie it 20 into Part 20. And also into an appropriate 21 relationship with environmental protection agency 22 requirements under Title 40.
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46 1 that the immediate health effects or prompt fatalities 2 would be less than five and ten million. And the risk 3 to, of latent health effects would be less than two 4 and one million years.
5 Again, based, that is the existing QHOs 6 just put out into words.
7 So, going on to Slide 19, just sensitive, 8 I received so much attention from stakeholders. We 9 have slides that we gave to the Subcommittee that 10 just, noting that many stakeholders did not believe 11 ALARA meant, a range of comments from ALARA shouldn't 12 apply to advance reactors down to, ALARA didn't need 13 to meet, in Part 53, because it was already addressed 14 in Part 20, to some proposing to keep it more or less 15 as we had proposed it.
16 Which is the same as it is provided in 17 Part 50. Specifically, although it's old, Appendix I 18 to Part 50.
19 So, our iteration has been, as we 20 discussed on the previous slide, to keep the ALARA 21 requirements in place. And looking forward, we did it 22 for occupational exposures, as well we for normal 23 effluence.
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47 1 the roll out of the preliminary language was on the 2 use of the QHOs. And again, a range of comments from, 3 don't include the QHOs, don't use the numerical 4 aspects of the QHOs and try to put it into more 5 general wording, to some who were in favor of 6 basically using them as we had proposed in the 7 preliminary language.
8 And our iteration has been to basically 9 keep them as we proposed in the first iteration. You 10 saw the language, we continued to refer to them as the 11 primary metric for unplanned events in the unlikely 12 and very unlikely event categories.
13 So, if we go then to Slide 21.
14 MEMBER DIMITRIJEVIC: Hi, this is Vesna 15 Dimitrijevic. I just want to make comment on your 16 previously slide. Where you said the QHO is a well-17 established measuring using risk-informed, I would 18 challenge that because the QHO are not directly ever 19 used in the risk-informed, just substantive measures.
20 MR. RECKLEY: Well --
21 MEMBER DIMITRIJEVIC: You know, CDF. And 22 no one ever in the application looks back to QHOs. I 23 mean, they are originally used to deny those CDF, but 24 they're based on the couple very significant 25 assumptions which have never been checked.
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48 1 So I would say the subsidy objectives are 2 well established, but not QHO. By any means.
3 MR. RECKLEY: Okay.
4 MEMBER DIMITRIJEVIC: They're based on the 5 conservatives and things like that. Like a 6 conservative change of like 30 percent to the aspect, 7 yes. Nobody ever goes back to QHOs.
8 MR. RECKLEY: And we'll have some slides 9 that Marty will talk about the QHOs and their 10 assessment. I will say more recently, before advance 11 reactors I worked in the area of the Fukushima 12 response, we used the QHOs.
13 When we were making determinations on 14 things like whether boiling water reactors should have 15 filters on the release, we were using the QHOs. When 16 we were looking at the assessment of whether we should 17 expedite the fuel, spent fuel transfer from poles to 18 casks, we used the QHOs.
19 So, yes, light water applicants have not 20 traditionally used the QHOs because, in large part, 21 surrogate measures have been developed. There have 22 been, in recent cases, the use of QHOs and decision 23 making.
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49 1 bit. Or we have backup slides when Marty talks about 2 probabilistic risk assessments.
3 So, I understand what you're saying. Not 4 disagreeing that for light water reactors the use of 5 surrogates, such as CDF and large release frequencies 6 have been used instead of QHOs. Marty can better 7 address the derivation of those surrogates. So, we'll 8 get to that in a few, in a few minutes.
9 MEMBER DIMITRIJEVIC: All right.
10 MR. RECKLEY: So, Slide 21, goes to 11 somewhat of a caution, if you will, that one of the 12 reasons we need a metric, and it would have to be, as 13 all the aspects of Part 53 has to be technology 14 inclusive, but we have to have a fairly high level 15 metric, but well-defined metric, within the safety 16 criteria, is because we are proposing to use those 17 metrics, again, throughout on how you procure 18 equipment on which quality assurance requirements 19 would be applicable.
20 Down into operations of how would one 21 define what the reliability targets are for equipment, 22 and that comes from the probabilistic risk assessment.
23 And a metric for that analysis, which we're currently 24 proposing to use the QHOs.
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50 1 metric, you make things like requiring an applicant to 2 define reliability targets for equipment that much 3 harder, if you don't have a metric to use for that 4 purpose. So that's all I wanted to say on Slide 21.
5 If we go to Slide 22, we had some 6 discussion during the Subcommittee meeting on the 7 safety functions. Basically we lay it out that the 8 primary function is the retention of radionuclides.
9 And then a requirement for additional 10 safety functions to be identified, and again, the 11 previous slides I had gone through on how to go 12 through that exercise. And largely what was done in 13 the late 1960s to develop the GDC was similar.
14 We gave examples of heat removal, heat 15 generation and chemical interactions. The ACRS 16 Subcommittee mentioned they thought reactivity should 17 be mentioned, and we'll commit to including that.
18 There was, I think as some other members 19 mentioned, we thought that was somewhat addressed but 20 maybe less clearly by saying heat generation. We 21 don't mind including reactivity as a specific example.
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51 1 box, one of the things that an applicant will need to 2 do is identify safety functions for every major 3 inventory.
4 And for some technologies, the other 5 inventories, like waste gas, can be comparable in 6 challenge to the reactor system itself. And so, under 7 the way we have it worded, hopefully they would need 8 to identify safety functions for that waste gas 9 system.
10 And reactivity would not, or maybe even 11 heat generation would be not as important. But they 12 would have to identify, for that waste gas system, 13 what are the safety functions needed to retain the 14 radionuclides.
15 As the secondary concern, and we'll talk 16 about this tomorrow when we talk about fusion, we have 17 said to the Commission that we would try to keep Part 18 53 so technology inclusive that it might address 19 fusion facilities.
20 And obviously the high level safety 21 function, retention of radionuclides, would be 22 applicable to even a fusion facility whereas 23 reactivity, as an example, or even post-operation heat 24 generation is less a concern. I won't say it's not a 25 concern, but it's less a concern than it is for a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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52 1 fission system.
2 So, that is part of the rationale for the 3 setup as we have it, for not listing the specific 4 functions other than the retention of radionuclides.
5 MEMBER MARCH-LEUBA: Hey, Bill, what are 6 you talking about? This is Jose.
7 I'm reading this as a rule. And you just 8 need an example of something that you might want to 9 consider to do what?
10 I mean, you might just well describe the 11 RFB, but what does it do?
12 MR. RECKLEY: Well, what it does is if you 13 go back to that, again, the first principles slide.
14 Any designer will have to identify how they are 15 planning to retain the radionuclides.
16 As Charlie mentioned, they all, what they 17 all have in common is there's a hazard. And that's 18 the radionuclides.
19 MEMBER MARCH-LEUBA: Okay. But what 20 you're saying is, Paragraph 53 to 30B is irrelevant.
21 And it doesn't tell me anything.
22 You just have to do, it basically says the 23 famous joke, when in doubt, refer to Paragraph A. So 24 either say something or don't say something. But what 25 you're saying doesn't say anything.
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53 1 I'll leave you with that concept. I mean, 2 rules have to be rules. I mean, they have to the law 3 and have to be well-defined.
4 When I read them, I need to know how to 5 follow it. I don't know how to follow this.
6 MR. RECKLEY: Well it's, again, it's 7 intended to give the designer enough flexibility to 8 say, to retain radionuclides, what functions do I 9 need. And so --
10 MEMBER MARCH-LEUBA: And that is the 11 function of a regulatory guide, not of the rule.
12 MR. RECKLEY: And we would expect that 13 there will be guidance in this area. And one way, 14 well, their actually already is guidance in this area, 15 in terms of Reg Guide 1.233 on the LMP, goes through 16 an exercise of identifying those safety functions. Of 17 laying out what would be needed in order to satisfy 18 Paragraph A on the retention of radionuclides.
19 MEMBER MARCH-LEUBA: You know what, if I 20 ask you, how do I satisfy Paragraph B, what do I have 21 to do?
22 Paragraph B is something I must do because 23 it's in the rule. What do I have to do to satisfy it?
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54 1 will be to say, what functions have you identified in 2 order to satisfy 53.230. And that would be, for 3 example, somewhat paralleled with the requirement we 4 currently have for light water reactors to address the 5 GDC and for non-light water reactors to define their 6 principle design criteria.
7 Those same things, as we talked about 8 before are, the identification of the safety functions 9 is part of that exercise.
10 MEMBER MARCH-LEUBA: Yes, I think I made 11 my point clear in that --
12 MR. RECKLEY: Okay.
13 MEMBER MARCH-LEUBA: -- this babble makes 14 no sense whatsoever.
15 MEMBER REMPE: So, I guess I have a 16 different perception, but make sure I understand 17 things correctly, Bill.
18 To me, they are going to go through. And 19 if they have a unique non-LWR, they will look at 20 things that could lead to radioactive, to release of 21 radioactive materials. And if they, for some reason, 22 if they have a chemical interaction and you have an 23 air ingress or a water ingress in a gas reactor that 24 can lead to radioactive material release, then that is 25 a safety function that will be identified as an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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55 1 additional safety function that must be done.
2 And then I look at C, and even though some 3 things are primary and some things are additional, all 4 of those things have to be met. And so, the fact that 5 it's a primary or an additional one doesn't mean the 6 regulatory is going to say, oh, it's only an 7 additional one. It gets the same attention as a 8 primary.
9 Am I understanding the intent of what the 10 words are here, Bill?
11 MR. RECKLEY: Yes. I think you probably 12 worded it better than I did, so thank you.
13 And the reason that it's constructed the 14 way we constructed it was, because it's to be 15 technology inclusive, how you do B might differ.
16 And again, I don't like to use it too 17 much, but ultimately if Part 53 is used for fission 18 energy systems, it will have a different set of safety 19 functions than fission reactors. Some of them will be 20 similar, but they'll be different than fission 21 reactors have.
22 But even within different designs, the 23 importance of something like chemical interactions 24 might differ. And so, anyway, I understand what 25 people are saying, we'll just take that as an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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56 1 observation.
2 23. So licensing basis events are, again, 3 I don't think the concept of having them has been very 4 controversial. There has been some discussion on 5 where this should be within Part 53.
6 But in general, what we're trying to 7 emphasis is that any designer needs to look at a range 8 of unplanned events from anticipated operational 9 occurrences down to very unlikely sequences. And 10 within LMP, if you want to go over to that 11 terminology, from AAOs to design basis events and the 12 lowest frequency events down into beyond design basis 13 event category.
14 So, going on then to Slide 24.
15 MEMBER REMPE: Oh, one more thing, Bill.
16 MR. RECKLEY: Oh.
17 MEMBER REMPE: Actually, back to 22. The 18 other thing I guess one I would raise to maybe address 19 some questions would be that, if an applicant didn't 20 identify chemical interactions and the staff reviewed 21 it and said you need to look at this because you could 22 have had a release, they would have to add that safety 23 function as part of the review process for Part 53, 24 right?
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57 1 raising questions. And quite possibly, it could lead 2 to them adding that as a safety function.
3 MEMBER REMPE: Again --
4 MR. RECKLEY: Yes.
5 MEMBER REMPE: -- that's I think something 6 that's important that might help with some of the 7 confusion about this. But again, I came from a 8 history and a prior career with an advance reactor 9 component and that's what they were concerned about.
10 But anyway, go ahead. Thank you. I'm 11 sorry to interrupt again.
12 MR. RECKLEY: Oh, no problem. Thank you.
13 MEMBER BLEY: Bill?
14 MR. RECKLEY: Yes.
15 MEMBER BLEY: I'm going to interrupt.
16 We're about halfway through your slides and I think 17 the next five or six are kind of really important and 18 things we didn't talk about in depth at the last 19 meeting. So I'm going to declare a break right now 20 and then we'll come back and finish those.
21 So Part C will probably go a little faster 22 than those. So, at this time I'm going to declare a 23 break. And we will recess until a quarter till the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
25 MR. RECKLEY: Okay, thank you.
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58 1 MEMBER BLEY: We're in recess.
2 (Whereupon, the above-entitled matter went 3 off the record at 3:27 a.m. and resumed at 3:45 a.m.)
4 MEMBER BLEY: At this time we will 5 continue with Bill Reckley's presentation. Bill?
6 MR. RECKLEY: Thank you, Dennis. So, one 7 of the things we wanted to talk about is the licensing 8 basis events.
9 There was some discussion at the 10 subcommittee meeting and some distinctions of how it's 11 done under Part 53 and the basis that we're getting 12 out of the licensing modernization project and maybe 13 how it was done traditionally, so the next few slides 14 are kind of a summary or a revisiting of the LMP and 15 the discussions we had with the ACRS during the 16 development of Reg Guide 1.233, SECY paper 19-0117.
17 I don't know the protocol, Dennis, so I'll 18 just offer up that I know there's new members. I also 19 know this is one of those topics that if you're not 20 exercising it, it's hard to keep in the forefront of 21 your mind.
22 So, if for new members or as kind of a 23 refresher for anybody, if there's a mechanism for us 24 to give presentations or whatever informal processes, 25 we're certainly willing to do that if there's an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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59 1 interest, and we can set that up if Derek or somebody 2 wants to just give me a call.
3 We have existing presentations we're using 4 for staff and in interactions with other regulators 5 like CNSC for example, so we have all of that on hand.
6 It's not really a burden for us to do.
7 MEMBER BLEY: Thank you very much. I 8 think that's something -- we'll talk about it.
9 MR. RECKLEY: Okay.
10 MEMBER BLEY: It's something I think Paul 11 might want to take advantage of, so go ahead, and this 12 is one more of those areas where the real language 13 just says we'll select them from this group, but it 14 doesn't really go into what was in your SECY, what's 15 in the LMP on exactly how you do that.
16 MR. RECKLEY: Right, so, but before, if we 17 go on then to slide 24, before getting into the LMP, 18 we just might want to revisit some of the ways it's 19 been done in the past.
20 And so I know it's a busy slide and most 21 of you are probably aware of the traditional 22 approaches. I just copied this out of the standard 23 ANS 51.1, the 1983 version.
24 Actually, for most operating reactors, it 25 was the previous version, the one that's kind of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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60 1 highlighted that talks about condition two, three, and 2 four events, basically the anticipated operational 3 occurrences and design basis accidents. This is also 4 discussed a little further in chapter, or, yeah, 5 chapter 15 of the standard review plan under section 6 15.0.
7 But basically it just lays out, and it's 8 similar for boiling water reactors and pressurized 9 water reactors, and kind of follows roughly a process 10 hazards kind of approach.
11 Consider what could make temperatures go 12 up and down. Consider what might make flow rates go 13 up or down. Consider what might make reactivity go up 14 or down, what might disturb the power distribution 15 within the core, what might lead to losses of 16 inventory.
17 And laid out basically in the earlier 18 versions was largely based on engineering judgment to 19 define the categories in terms of anticipated events 20 or events that were not considered likely to happen or 21 conditioned for the design basis accident conditions.
22 The 1983, by the development of the 1983 23 standard, you can actually see that frequencies were 24 being considered more specifically in both the 25 categorization and also the little box is basically a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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61 1 frequency consequence curve.
2 So, by the 1980s, the earlier versions 3 were starting to introduce or including more of a 4 frequency component. Now, very -- I don't think any 5 reactor was actually referencing the '83 standard. By 6 that time, we weren't licensing plants anymore.
7 So, that kind of just lays out the 8 background for the light water reactors. If we go 9 onto slide 25, I'll go through a few slides that 10 basically gives a similar process as it was developed 11 under LMP, one difference being instead of using the 12 -- it may be process hazard oriented terminology of 13 consider what can make flow rates go up and down.
14 It basically is actually looking at event 15 sequences from the PRA and looking at a particular 16 component and failing it one way or another. In the 17 end, it's similar.
18 So, depending on how you want to approach 19 it, you can either highlight the differences or you 20 can actually highlight the similarities between even 21 the historical approach and the LMP.
22 MEMBER BLEY: Bill?
23 MR. RECKLEY: Yes, Dennis?
24 MEMBER BLEY: I actually had two things.
25 The first is I agree with what you just said, but the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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62 1 underlying PRA --
2 (Telephonic interference.)
3 MEMBER REMPE: Dennis, I think we lost 4 you.
5 MR. RECKLEY: Okay.
6 CHAIR SUNSERI: Yeah, Dennis, this is 7 Matt. If you -- we can't hear you if you're talking.
8 MR. RECKLEY: Okay, what I might do, and 9 Dennis has had some problems, I know, from the 10 subcommittee meeting, so maybe I'll go on, and then 11 when he comes back, we can pick up. Is that okay?
12 (Simultaneous speaking.)
13 CHAIR SUNSERI: He's asked me to carry on 14 if he --
15 MR. RECKLEY: Okay.
16 CHAIR SUNSERI: -- drops off, so go ahead, 17 Bill, carry on.
18 MR. RECKLEY: Okay, and we can revisit 19 when he reconnects with the point that Dr. Bley was 20 going to make.
21 So, basically for the LMP, the event 22 selection is again taking the event sequences from the 23 results of the probabilistic risk assessment and 24 plotting them in terms of frequency and consequence 25 onto this figure, and then as we get into more of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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63 1 discussion, ultimately looking to the margins that 2 exist between those events and the target figure, 3 which is the figure in blue.
4 And if we go down to slide 26, this is a 5 slide -- again, I'm using the MHTGR, trying to use it 6 so we can maintain some consistency between what we're 7 talking about, but this is one of the tabletop 8 exercises done for the LMP.
9 It was actually done for X-energy XE-100 10 design, but where they were at this time. This was 11 four years ago. They were looking and largely 12 borrowed by the MHTGR PRA and event assessments, and 13 you can see in purple all of the event sequences that 14 they had identified in the various categories.
15 And then as we get into more of the 16 discussion, actually the red dots, if you can see them 17 in the design basis event region, are event sequences 18 that contribute to the identification of a design 19 basis accident, and we'll get into that discussion 20 when we talk about the DBEs.
21 So, it does just show the number of events 22 and the number of sequences that are being looked at 23 from the probabilistic risk assessments, and actually 24 even this, the number, if you plotted all of the 25 sequences that were actually run, there would even be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 1 more dots on here.
2 They do group them into what's called 3 event families based on similarities of how the 4 progression and the end state of the transient, so 5 this is actually a plot of families, not necessarily 6 individual sequences.
7 But, so if we go on then to slide 27, this 8 just lays out again the categories of events.
9 Anticipated operational occurrences are basically 10 those that go down to a frequency per plant year of 11 ten to the minus two, where a plant year is any number 12 of modules that might be affected, any number of 13 inventories that might be affected within a plant.
14 DBEs are between ten to the minus two and 15 ten to the minus four, and beyond design basis events 16 below ten to the minus four down to five times ten to 17 the minus seven, and then importantly, the methodology 18 includes the assessment of uncertainties and the 19 requirement to really look at that, and if the 95th or 20 the fifth percentile in an uncertainty assessment puts 21 you across the band, then you look at it in both 22 categories.
23 So, the other thing that I'll just point 24 out is, and this, we've introduced some confusion. In 25 an attempt to not use exactly the terminology that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 1 used in the LMP or to give the impression that we were 2 requiring they use LMP, we're introducing different 3 terms, but the terms have the same meaning, like 4 instead of beyond design basis events, you might have 5 noticed we call them very unlikely events, so, but 6 within the overall construct, they're the same.
7 One last point on this, and I know I'm 8 going pretty quickly through what can take many 9 minutes to discuss, the other thing that's looked at 10 under this is an assessment when you're looking at all 11 of these event sequences against the cumulative risk 12 metric, and again, that is proposed to be the QHOs in 13 our particular example.
14 So, the other aspect, if we go to slide 28 15 --
16 MEMBER MARCH-LEUBA: Wait a minute, Bill, 17 go back, go back. I wanted to make a comment on the 18 record. How did you, I mean, how do you address in 19 this methodology the known unknowns, which is what 20 they call the completeness of the PRA? How do you 21 know you selected all of the events at the front of 22 that, not just the ones you thought about, but all of 23 them?
24 MR. RECKLEY: Well, I think I'll let Marty 25 get into that discussion a little more, but it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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66 1 basically goes through the methodologies that you use 2 to make sure you address everything that can break.
3 MEMBER MARCH-LEUBA: No, that goes against 4 the scientific method. You cannot prove a negative.
5 You cannot say I -- you can say of everything I 6 looked, this is how it turns out to be, but there 7 might be something else I didn't understand.
8 And with light water, large light water 9 reactors, we have 60 years of experience. Basically, 10 almost everything that could happen has already 11 happened.
12 With these large reactors, we don't have 13 any experience and we have designers that want to 14 expedite things. They don't, you know, have as much 15 money. They cannot spend 20 years designing a 16 reactor. They have to do it in two.
17 The completeness of the set of events is 18 crucial. It's crucial, and as I keep telling you, I 19 mean, this is not a hypothetical, okay? You can 20 forget the most limiting events simply because it 21 doesn't fit in what happened for the last 60 years on 22 light water reactors.
23 And I don't see any emphasis on the rule 24 or in your thinking on the review of the staff to 25 understand how complete is that set of events.
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67 1 MR. RECKLEY: Okay, let's --
2 MEMBER MARCH-LEUBA: And that's a clear --
3 I mean, you say you're premising the whole Part 53 on 4 the fact that I can calculate the risk, and that's a 5 non-scientific statement, period, over and out.
6 MR. RECKLEY: Okay.
7 CHAIR SUNSERI: Bill, this is Matt. I 8 just want to let you know Dennis is back on, so.
9 MR. RECKLEY: Okay, let's revisit that 10 point when we get to the PRA discussion, and I think 11 Marty will, I think either in his slides or in the 12 backup slides, go through the methodologies, but let 13 me defer that. And Dennis, if you had a point before 14 you dropped off?
15 MEMBER BLEY: It's an important point Jose 16 raises, but it's not a point about PRA. It's about 17 safety analysis. It would apply whether we were doing 18 PRA or the other kinds of events, and the things that 19 tend to dominate risks aren't things that we would 20 have seen in 60 years.
21 They are things that -- we haven't seen 22 everything. We're going to see some more things. And 23 so I'll be happy when Marty gets to this and talks 24 about it, but it's not strictly a PRA issue at all.
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68 1 revisit this. Let me go on then to slide 29, 28, 2 thank you.
3 So, one of the things that comes out of 4 the process as we talked about are what are the 5 required safety functions, things like heat removal 6 and reactivity, and for those, those are the functions 7 that have the potential to make you exceed the 8 frequency consequence targets, and in particular in 9 our example for the first tier safety criteria, the 10 potential to exceed the 25 rem reference values.
11 That is what then goes into the 12 determination of safety-related equipment because for 13 every required safety function, you're required to 14 have safety-related equipment in order to do what's on 15 the next slide, the design basis accident, and 16 demonstrate that using only safety-related equipment, 17 you don't exceed the referenced values.
18 So, if we go onto slide 29, again, just 19 coming back to the MHTGR example, they've done this 20 exercise. They've identified the required safety 21 functions, and then you go down into slide 30.
22 They would use only safety-related 23 equipment to perform those functions in the DBA. So, 24 they're going to have a safety-related reactivity 25 control system, a safety-related heat removal system.
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69 1 Many of the non-light water reactors are using reactor 2 cavity cooling or reactor vessel direct cooling 3 systems.
4 But in any case, you have a safety-related 5 system for those required safety functions of needing 6 to bring down the heat generation through reactivity 7 control and to remove that heat through a decay heat 8 removal system, so the DBAs are derived from the PRA 9 sequences and then are looked at again only using 10 safety-related equipment.
11 So, this is the LMP approach. It's also 12 the approach that's reflected in Part 53 to have both 13 a PRA, or as we'll talk about, another systematic 14 assessment, and to keep a fairly deterministic DBA, 15 traditional safety-related equipment as a kind of a 16 backstop for the plants.
17 So, if we go down then to slide 31, the 18 other couple sections that remained in Subpart B, the 19 safety criteria, is the defense in depth 53 250.
20 Again, we didn't make any major changes in the second 21 iteration.
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70 1 given special consideration, but no single engineered 2 design feature could be relied on to meet the safety 3 criteria in 53 220(b), which again is meeting the 4 QHOs.
5 MEMBER BLEY: Bill?
6 MR. RECKLEY: Yes?
7 MEMBER BLEY: That's a nice distinction, 8 but it implies if you have an inherent feature, you'd 9 be happy with a single one, and depending on what you 10 -- they can be degraded as well, so it seems odd to 11 suggest engineered design features to me.
12 MR. RECKLEY: And we're going to have to 13 define some of these terms. Engineered design feature 14 would include a passive system, so those can be 15 degraded. What we're tend --
16 I mean, we're still developing this and 17 engaging stakeholders on the terminology, but when we 18 use the word inherent, it is something that doesn't 19 require something even like natural circulation. So, 20 it's not --
21 MEMBER BLEY: Even if it's coming from the 22 physics, which I think is what you're saying.
23 MR. RECKLEY: Right.
24 MEMBER BLEY: You really got to be careful 25 and make sure, one, you know all of the physics that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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71 1 might apply, and two, that nothing outside can 2 interfere with it, so.
3 MR. RECKLEY: I 100 percent agree and 4 that's -- and we're not trying to say it would be 5 easy. So, if you have an inherent feature, you're 6 right.
7 What we mean by that is it's the physics, 8 but the physics has to be maintained over the life of 9 the facility, so that means the physics couldn't be 10 changed by irradiation or other environmental factors.
11 It means the inherent feature is present within the 12 bounds of what the plant's going to be operating 13 under.
14 So, no easy task to show that the inherent 15 feature can be relied on. They're going to have to do 16 the science, the testing, and all of that to 17 demonstrate that that inherent feature could be relied 18 on, so.
19 MEMBER BLEY: Then you have plenty of 20 external things like fires, severe earthquakes.
21 MR. RECKLEY: Right, so --
22 (Simultaneous speaking.)
23 MEMBER BLEY: -- all of those things.
24 MR. RECKLEY: We agree 100 percent. We 25 were just trying to give some room that if there is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 1 such an engineered, I mean, if there is such an 2 inherent feature and it can be proven, then that would 3 be a basis to at least evaluate not requiring 4 additional defense in depth measures, but no easy 5 task. I'd agree with you there.
6 MEMBER PETTI: Bill?
7 MR. RECKLEY: Mm-hmm?
8 MEMBER PETTI: Does this mean as written 9 that if one wanted redundancy and backup, you could 10 have one engineered system and one inherent system?
11 MR. RECKLEY: That would be one way to 12 address the potential uncertainties with the inherent 13 characteristic that Dennis just mentioned, yeah.
14 MEMBER PETTI: Right, I can see some 15 cases, some inherent functions where you can back it 16 up. I can see others that it's harder to back up, for 17 instance, molten salt. The fission product retention 18 in that salt, I don't know you'd get an engineered 19 system there.
20 I can -- probably an engineered system to 21 make sure the temperature doesn't get outside some 22 bound that would invalidate its ability to hold 23 fission products or something, but okay, thanks.
24 MEMBER KIRCHNER: Bill, this is Walt.
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73 1 have to modify it with engineered? Why not just no 2 single design feature, whether it's inherent, or 3 passive, or engineered, or -- it's a design feature, 4 something maybe as simple as a negative temperature 5 coefficient or neutron leakage for reactivity control.
6 But again, as you pointed out and Dennis 7 did in his examples, those things can be affected 8 throughout the life of the plant because of upset 9 conditions and so on and so forth.
10 You know, like a fast reactor that depends 11 on leakage, well, you might not have that performance 12 characteristic under all conditions, et cetera. Why 13 not just leave it at single design feature and not 14 have to split hairs over whether it's engineered, 15 passive, or inherent?
16 MR. RECKLEY: We were -- I mean, one of 17 the reasons is you have to go back and see how this is 18 actually being used to assess individual event 19 sequences, right, individual events.
20 And so whereas what we're talking about up 21 to this point, I tend to agree with everyone that when 22 you're talking at an overall plant, that is actually 23 the way it would most likely play out, but when you're 24 looking at an event, at a particular event, we're just 25 saying if it could be proven that that event could be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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74 1 addressed by an inherent feature, we were 2 acknowledging it's a challenge, but we were trying to 3 say that that sequence then would not need to be 4 backed up with an additional design feature, or may 5 not have to be.
6 So, again, it was one of the comments that 7 we had gotten. Some of the designers had felt 8 strongly they could justify the inherent features, and 9 so as a compromise, this is what we're proposing, but 10 I guess that's all the explanation I can give.
11 MS. VALLIERE: Hey, Bill? I might add 12 just to jog the members' memories that when we 13 presented on key guidance documents that need to be 14 developed to support Part 53, you'll find I think in 15 that list that guidance on inherent characteristics 16 was one of the items identified as needing guidance to 17 support Part 53.
18 MR. RECKLEY: Thank you, Nan. Okay, so if 19 we can go onto slide 32, the last section within 20 Subpart B on the overall objectives and safety 21 criteria is the need to protect plant workers. We 22 largely do this by referencing back to Part 20, and I 23 don't think there was much controversy to that, at 24 least in the discussions with the subcommittee.
25 Going on then to the next section and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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75 1 next subpart under design and analysis -- well, maybe 2 I'll stop there and say is there any questions or 3 discussion on Subpart B?
4 MEMBER BLEY: Bill?
5 MR. RECKLEY: Yes?
6 MEMBER BLEY: I was trying to make a point 7 before I lost the internet.
8 MR. RECKLEY: Yes.
9 MEMBER BLEY: I don't know if you heard 10 me.
11 MR. RECKLEY: Only the very first 12 fragments, so, yes, if you could just repeat the two 13 points?
14 MEMBER BLEY: There were two things I 15 wanted to mention. With respect to your slide number 16 24, which is kind of nice, but the first note is by 17 the time you have this kind of layout, you'd have 30 18 years' experience working almost exclusively with 19 expert judgment to dream about what are the things 20 that could go wrong and how do we consider them, so 21 this was a real evolution by the time you got here.
22 And the other, I think I was just talking 23 about whether you're doing this traditional approach 24 to define your errors and design basis accidents that 25 you're going to analyze in the traditional way or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 1 whether you're looking for initiating events in some 2 areas in the PRA, it's the same process.
3 You've got to find them before you know 4 what they are and that's a place where this time it 5 looks pretty coherent, but that was after 20 to 30 6 years of trying to describe what these things ought to 7 be and it was unique to LWRs.
8 So, this idea that you need a way to look 9 for these events, especially for new technologies 10 where we haven't been working on them for decades, 11 that's where the guidance for people is very sparse.
12 There isn't --
13 (Telephonic interference.)
14 MEMBER BLEY: Please continue with your 15 next set of slides.
16 MR. RECKLEY: Okay, thank you, Dennis, and 17 I'll also mention that, you know, one of the ones with 18 the least experience is molten salts, and there are a 19 couple of reports out of Oak Ridge going through an 20 exercise similar to ANS-53.1.
21 There's also a good EPRI report that was 22 supported in part by DOE that talks about how to do a 23 process hazard analysis for molten salt systems, which 24 are similar.
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77 1 with molten salt reactors, but chemical systems have 2 been used in process hazards analysis for a long time, 3 so they use that exercise.
4 And then the EPRI report talks about, as 5 Marty will go into the PRA discussion, also how to 6 inform or to use the process hazards as a starting 7 point for what ultimately would go into the PRA, but 8 we'll talk about that a little more under Subpart C 9 under the analysis.
10 So, yeah, if we go onto slide 34, again, 11 the layout of Subpart C follows the chevrons we talked 12 about earlier, the design criteria, the safety 13 functions in Subpart B.
14 Then they progress down into Subpart C 15 where the first section is on design features, and 16 then the second section, second and third sections are 17 how do you define the functional design criteria to 18 meet the first tier.
19 That's the safety-related design basis 20 accident tier, and then the second tier, which is the 21 more risk-informed approach coming out of the PRA, the 22 beyond design basis events and so forth, and then how 23 you get down into some additional design requirements 24 we'll talk about, and then really we want to spend 25 some time talking about the role of the PRA under the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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78 1 analysis section.
2 So, going through the first couple of 3 sections, 35, I'll turn it over to Marty. One of the 4 things that we wanted to talk about because it had 5 come up in the subcommittee meeting was single failure 6 versus the PRA probabilistic and reliability approach, 7 so, Marty?
8 MR. STUTZKE: Yeah, good afternoon. I'm 9 Marty Stutzke, the senior technical advisor for 10 probabilistic risk assessment in the division of 11 advanced reactors and non-power production and 12 utilization facilities.
13 And as Bill had said before, we wanted to 14 talk about the fact that Part 53 would allow the 15 single failure criteria to be replaced with a 16 reliability criteria.
17 This had been mentioned in Reg Guide 1.233 18 as approved by the SRM, the SECY-190117, to allow us 19 to do this, as well as using probabilistic evaluation 20 to select events, some things like that.
21 A little bit prior to that in a different 22 context was the staff had approached the Commission in 23 SECY-19036 about the NuScale ECCF systems, 24 specifically the inadvertent actuation block valves, 25 and whether single failure criteria should apply to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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79 1 those valves specifically, and the Commission came 2 back and told us to apply risk-informed principles 3 when you don't need the strict deterministic criteria 4 such as the single failure criteria.
5 MEMBER BLEY: Marty?
6 MR. STUTZKE: Yes?
7 MEMBER BLEY: This may be more for Bill.
8 When you're all talking about the single failure 9 criteria as applied to a system, and a big system that 10 has a safety function, under the single failure 11 criteria, it has to be able to withstand any single 12 failure without a loss of function.
13 There's another aspect of single failure 14 that Bill was talking about earlier, and that is when 15 you do the equivalent of the Chapter 15 analysis 16 deterministically with only safety grade equipment 17 operating, you assume for each system the most 18 challenging single failure.
19 That is still part of Part 53 as I 20 understand Bill's earlier explanation. Is that 21 correct, Bill?
22 MR. RECKLEY: Actually not. The defense 23 in depth measures that we talked about would require 24 that you have additional measures, but the difference 25 between what we're proposing and the traditional NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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80 1 single failure criteria is that we wouldn't require 2 for the DBA a specific additional single failure that 3 has led traditionally to two trains.
4 MEMBER MARCH-LEUBA: It would be perfectly 5 okay with you to have a single safety protection 6 system, a one channel protection system because you 7 would have to assume a single failure?
8 MR. RECKLEY: One train, yeah, one train.
9 MEMBER MARCH-LEUBA: So, your I&C will not 10 only not have diversity, it won't even have 11 redundancy?
12 MR. RECKLEY: For the assessment of the 13 DBA. The diversity --
14 (Simultaneous speaking.)
15 MR. RECKLEY: For the diversity, and the 16 redundancy, and so forth comes largely in repeating 17 that in a non-safety related system most likely for 18 the other event sequences.
19 MEMBER MARCH-LEUBA: So, you will have one 20 safety grade I&C channel and say three non-safety 21 grade channels, trains? That would be perfectly okay?
22 MR. RECKLEY: I'll be honest. I haven't 23 seen the application of this down to the I&C channel.
24 I'm mechanical oriented, so --
25 MEMBER MARCH-LEUBA: Okay.
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81 1 MR. RECKLEY: -- I've seen it on the 2 mechanical --
3 MEMBER MARCH-LEUBA: You're a mechanical 4 guy. Would it be okay to have one single safety 5 relief valve to protect for the SME safety code?
6 MR. RECKLEY: For the DBA, yeah.
7 MEMBER MARCH-LEUBA: Yeah, so only one 8 safety relief valve will be okay for you --
9 MR. RECKLEY: But keep in mind --
10 MEMBER MARCH-LEUBA: -- to protect this 11 against other pressure?
12 MR. RECKLEY: For the DBA. Because you 13 have to analyze the other events and meet the defense 14 in depth requirement, you will have more than one.
15 (Simultaneous speaking.)
16 MEMBER MARCH-LEUBA: But it will not be 17 safety related.
18 MR. RECKLEY: It may not be safety 19 related.
20 MEMBER MARCH-LEUBA: So, only one safety-21 related SRV, only one safety-related protection 22 channel, only one control, okay, that's fantastic, 23 man. You're making my day.
24 MR. RECKLEY: Keep in mind that you're 25 talking about what is needed to protect against an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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82 1 individual event sequence, not what would be found 2 acceptable for the overall plant design, because you 3 need to bring in the other event categories, and the 4 defense in depth requirement, and other --
5 MEMBER MARCH-LEUBA: The only event 6 categories, the ones that don't have AOOs don't give 7 you safety-related components. Are you saying that 8 we're going to create non-safety grade, some 9 additional control protection system channels and 10 trains, non-safety grade SRVs we're going to give them 11 credit for? They're not in tech specs and do not 12 exist, but we grade them? Okay, guys, you know how I 13 feel about this thing. This is lunacy.
14 MR. RECKLEY: Okay --
15 MEMBER BLEY: Bill? This is -- yeah, when 16 you do the DBA analysis, which Part 53 calls 17 deterministic, that's fine and conservative, but you 18 assume everything's working. You're not doing 19 reliability accounting for the chance of failures.
20 You're assuming everything works, so it's a different 21 kind of analysis that we did before.
22 Now, I will agree with you if you've done 23 the PRA right, you've looked at the overall risk and 24 the chance that things fail, but if you do that and 25 you come up with those licensing basis events which NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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83 1 are out of the PRA, those are pretty reasonable, but 2 then when you say I'm going to define a DBA as one of 3 those and I'm going to analyze it in the traditional 4 deterministic conservative way, you're not doing that.
5 Now, there are good arguments about why 6 you might not want to define DBAs, just stay with the 7 overall PRA, but I don't see what you gain at all by 8 defining DBAs and then applying thermal hydraulics to 9 it. I don't get it.
10 MR. RECKLEY: The notion is that you'll 11 have, at least for the required safety functions, 12 you'll have at least one safety-related way to meet 13 that function. So, in reality, you have multiple, but 14 at least one of those paths will include only safety-15 related equipment, so.
16 MEMBER BLEY: But if we go back to 17 thinking about an LWR, when we have, say, three pumps 18 of safety injection and you go buy one that's safety 19 grade and two that aren't, you're probably going to 20 buy the same pumps. I'm not sure what we're picking 21 up here.
22 MEMBER REMPE: Well, would we say the 23 maintenance might be less for the non-safety grade 24 ones?
25 MEMBER BLEY: It might be nonexistent.
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84 1 MEMBER REMPE: Right, so --
2 MEMBER BLEY: And it can't be, and you 3 can't just have one because when you do that PRA with 4 its embedded systems analysis, you cannot get 5 sufficient enough reliability out of the system if you 6 don't have maintenance and if you don't have 7 redundancy in the systems.
8 You can't approach anything like returns 9 of reliability we need in our systems to protect the 10 design. I guess where I'm -- it sounds like the only 11 thing doing this defining of the DBA that does 12 anything is that the main one of them is safety grade.
13 They're all going to have to be under tech specs or 14 you can't get the maintenance contributions on 15 reliability well enough.
16 MR. RECKLEY: And we'll get there when you 17 see the operating controls we set. There will be 18 reliability programs for -- let's take your example 19 and there's three ways to remove heat.
20 Yes, the exercise is one of those ways, if 21 it's serving a required safety function, it will be 22 safety related. They will all have, if they're risk 23 significant, they'll all have reliability controls on 24 them.
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85 1 is the safety related would have the control under 2 tech specs and the other two would have their controls 3 under a reliability assurance program, not in tech 4 specs, but in another required by regulation program, 5 which is the reliability assurance program, so.
6 MEMBER PETTI: So, Bill, does that -- I 7 mean, I'm trying to understand does that change 8 anything really on the ground that there's 9 requirements, but they're coming through two different 10 pathways if you will in terms of what you do with the 11 systems on the ground?
12 MR. RECKLEY: The thought is there would 13 not be that much difference on the ground. From a 14 regulatory perspective, the tech specs will have the 15 traditional action statements and so forth, whereas 16 the others would come more under a licensee-defined 17 program, so a little more flexibility in the non-18 safety related, non-tech spec.
19 In terms of the actual equipment like 20 you're suggesting, probably not that much difference, 21 but in the regulatory treatment, some difference.
22 MEMBER PETTI: Okay.
23 MR. RECKLEY: So, I'm sorry about --
24 MEMBER REMPE: When I think about like 25 crud that was deposited on the vessel head nozzles, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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86 1 how often they have to be inspected could change, 2 which could affect the performance, right?
3 And so I'm trying to think of examples 4 with real operating plants where you had to do stuff 5 and there was an inspector who was verifying it was 6 done. I mean, yeah, it would save a lot of money for 7 the licensee, but I'm wondering does that mean we're 8 really depending on one?
9 MR. RECKLEY: Yes, some of this, I think, 10 and I hate to say it, but I think it will be more 11 clear when we look at the operations requirements, and 12 that will be next month, so.
13 MEMBER KIRCHNER: Bill, this is Walt. I 14 guess I'm -- going back to my colleague Jose's 15 comments, let's just pick something, a reactor 16 protection system.
17 To only have one channel, to me, violates 18 the whole philosophy of defense in depth, one safety-19 grade channel for detection of, let's say, over power, 20 you know, high flux calibrated in terms of power, so 21 a power trip.
22 MR. RECKLEY: And if we could, Walt, just 23 because I'm not as familiar with the I&C side and the 24 fact that when you get into I&C, even in the safety-25 related functions, you've going to have multiple NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 channels.
2 Those channels might actually all be 3 safety-related channels because they're looking at 4 different quadrants in the core. They're looking at 5 different loops in the coolant system and so forth.
6 So, I prefer not to focus in on an I&C 7 channel, but if we look at, let's say, a heat removal 8 system and take reactor cavity cooling as the safety-9 related system, all we're saying is there wouldn't be 10 necessarily two trains of reactor cavity cooling, but 11 reactor cavity cooling is not the only heat removal 12 system you have.
13 In fact, it might be the fourth or fifth 14 heat removal system that you have, but it might turn 15 out to be the safety-related system you have for heat 16 removal.
17 MEMBER KIRCHNER: Well, let's take it on 18 the mechanical side, just kind of rhetorically, your 19 nice chart of the layered fission product or 20 radionuclide barriers.
21 So, and maybe the first one is the 22 equivalent of the, of a fuel form, or the first one is 23 for a liquid fueled system is probably that primary 24 envelope.
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88 1 first line of defense, one would think that you would 2 need a safety grade second line of defense or, yeah, 3 safety related, I'm sorry, equipment, you know, 4 qualified to be in concert with your defense-in-depth 5 overlying not philosophy now but objectives in terms 6 of --
7 MR. RECKLEY: Right.
8 MEMBER KIRCHNER: -- system performance.
9 MR. RECKLEY: And I think we're probably 10 in agreement other than the safety classification of 11 your backup.
12 The fact is you would be required to have 13 a backup. But the backup, depending on the assessment 14 that you're doing, the backup would very likely be a 15 non-safety related backup.
16 And that's not dramatically different than 17 what we accepted on some of the passive light water 18 reactor designs. But it -- there still would be a 19 backup. There still is defense-in-depth. You're not 20 totally relying on one layer as you're suggesting.
21 It's just that, because of the way we've 22 categorized the events, the design-basis accident, 23 you're going to credit the safety related one. You're 24 going to ignore the non-safety related ones that are 25 actually providing that backup.
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89 1 And that way you're assured, again, that 2 you have at least one safety related way to carry 3 that, to make sure that you don't exceed the 25 rem 4 first tier safety criteria.
5 (Simultaneous speaking.)
6 MEMBER KIRCHNER: I understand what you're 7 saying. I just, you know, I've said this too many 8 times in the past. But I'll say it one more time.
9 I kind of look at this and say, well, does 10 this provide an equivalent level of protection in the 11 public's eye, I mean, because that's, you know, if 12 that defense-in-depth, the second barrier now, is not 13 safety related, do you convince the public that you've 14 provided an adequate, a comparable level of safety to 15 the existing fleet. And I don't know.
16 It strikes me that the public, looking at 17 this not knowing the nuances of an in-depth PRA, et 18 cetera, et cetera, might not be convinced.
19 MEMBER HALNON: Bill, do you ever see a 20 situation where the backup would be non-safety related 21 but important enough to be tested by a tech spec 22 surveillance?
23 MR. RECKLEY: We're still developing the 24 requirements under Subpart F. That question comes 25 down really to the fourth criteria and under 50.36 for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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90 1 including limiting condition for operation and a tech 2 spec surveillance for risk significant SSCs.
3 And since it's out, I'll tell you, our 4 first draft, our first iteration of the language says 5 that that's addressed through the reliability 6 assurance programs for the non-safety related 7 equipment and not included in tech specs. That's our 8 first iteration.
9 MEMBER KIRCHNER: But what happens, Bill, 10 when the PRA results are used such that that second 11 backstop or second line of defense after the safety 12 related equipment has been assumed to fail or does 13 fail and it's not on the D-RAP?
14 And I don't want to go into actual 15 details, but when we have instances where in the 16 recent review the two obvious systems to recover or 17 provide that backup didn't make the D-RAP list.
18 MR. RECKLEY: Well, again, well, the 19 assumption -- and I'll let Marty get back into the 20 slides here. But the -- if it's shown to be either 21 risk significant because of the PRA results or it's 22 required to meet the defense-in-depth measure, under 23 what we would propose under Part 53, it would be in 24 the equivalent of D-RAP. That's one of the criteria 25 for being there.
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91 1 MEMBER KIRCHNER: But let me pursue it a 2 little further then. So, okay, we obviate the need 3 for single failure criteria. We do define -- this 4 morning we heard about fire safety. I'll use their 5 terminology, a success path to fall below, in that DBA 6 analysis, fall below the dose, the safety criteria as 7 expressed in terms of dose at the exclusionary 8 boundary or LPZ.
9 Wouldn't, in that case, wouldn't that 10 second line of defense then have to be covered by the 11 D-RAP, or as Greg was saying, Greg triggered me on 12 this, that, wouldn't then that have to be somehow in 13 the tech specs at least for the DBAs?
14 MR. RECKLEY: Again, and we're jumping 15 ahead a month to look at what's in tech specs. But, 16 yeah, if it's required to address the DBA under the 17 Part 53 proposal we just released, then it's required 18 to be in tech specs.
19 MEMBER KIRCHNER: Okay. All right. Thank 20 you.
21 CHAIR SUNSERI: Hey, Bill, this is Matt.
22 Vesna has her hand up. You might want to call on her.
23 MR. RECKLEY: Please, Vesna.
24 MEMBER DIMITRIJEVIC: Yes, hi. So my 25 question is to Marty. Marty, are we discussing, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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92 1 because now I'm confused with all of this.
2 Are we discussing here, so action of the 3 licensing basis events and application to Chapter 15, 4 or we are discussing safety classification, because I 5 have a question for later on safety classification, 6 but suddenly we are discussing safety classification 7 here. And I didn't see too much about safety 8 classification in your documents.
9 So are we discussing here safety 10 classification of SSCs, or we are discussing selection 11 of licensing basis events? Those are two separated 12 things.
13 I mean, you say in selection of licensing 14 basis events, credit only safety equipment. But where 15 is the safety classification and how it's determined 16 that's not discussed.
17 MR. RECKLEY: Yeah, no, I might have gone 18 through the slides quickly. If you can flip back up 19 a couple to -- and the topics you mentioned are all 20 interrelated.
21 If you go back to 27, so this is the 22 selection of the licensing basis events by looking at 23 the PRA sequences and putting them in these categories 24 --
25 MEMBER DIMITRIJEVIC: Yeah.
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93 1 MR. RECKLEY: -- based on the frequency.
2 Then if you go to the next slide 28, you're looking at 3 those and saying, out of those event sequences, if I 4 take a piece of a -- what are the required safety 5 functions?
6 So, if I don't have any system to remove 7 heat, what would happen to the event sequence that I 8 just plotted in the design-basis event category?
9 Let's say I have three ways to remove 10 heat. And that's -- so that puts me in the design-11 basis event category, because probably at least two of 12 the three is going to work on any given sequence. But 13 if I take away all heat removal, what happens?
14 And that's what that arrow is showing. If 15 I take away all heat removal, then I'm likely in this 16 example to exceed the 25 rem number. I'm going to 17 exceed that frequency consequence target figure.
18 That means that's a required safety 19 function, because without it I won't pass the 20 criteria, the first tier criteria.
21 So now that goes into safety 22 classification. Given I have identified that as a 23 required safety function, I need at least one system 24 to perform, one safety related system to satisfy that 25 function, to perform that function.
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94 1 So one of, in the example, one of the 2 three heat removal systems is going to get picked as 3 a safety related system.
4 MEMBER DIMITRIJEVIC: Okay. So let's 5 start with this selection. This selection of event 6 sequences where they actually, the end state is not 7 determined yet. So it could be an initiating event, 8 everything successful, for example. That could be one 9 of the sequences, you know.
10 So let's say this sequence, if you want to 11 determine this lease in this category but you're only 12 crediting safety equipment, so we have here a question 13 of the chicken and egg.
14 I mean, how do you select sequences if you 15 don't know what the safety equipment or, I mean, you 16 know, you can see how this is all -- and somebody was 17 proposing that maybe we go as an example to this, I 18 think it was Bill, to this licensing basis event 19 selection, you know.
20 But we can actually do this as a tabletop 21 to see how that will work, because obviously these 22 things are so interconnected. It's not clear at all 23 how that will work in practice.
24 MR. RECKLEY: Right. And --
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95 1 me.
2 MR. RECKLEY: Yeah, and there's been 3 various tabletops done. And so we could revisit that, 4 as I mentioned to Dennis, through the LMP briefing or 5 whatever.
6 MEMBER DIMITRIJEVIC: I mean, examples in 7 that Lorad (phonetic), you know, which has examples 8 for PWR and PWR, the 1860, in those examples it's 9 already known, because they rely on the existing light 10 water reactor.
11 Like, for example, if you just look at 12 heat removal, the main feedwater, and feed and bleed, 13 they're non-safety functions. But here when you are 14 playing with, and then they're only looking in, you 15 know, absolutely feedwater removal drains.
16 But if you are looking in the new designs 17 and you don't really know what is happening, then you 18 don't know how to select those sequences in the basis.
19 So this is a -- you know, like looking at examples 20 will help a lot in these cases.
21 MR. RECKLEY: Okay. So, Marty --
22 MEMBER PETTI: So, Bill, just, let me make 23 sure I understand, because this is kind of moved 24 around.
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96 1 know, MHTGR, the regular cooling system was one way to 2 remove heat. It's clearly safety related. Reactor 3 cavity cooling system was a passive system. And it 4 was also safety related.
5 But there was a third cooling system that 6 many people don't know about called the shutdown 7 cooling system. It was not safety related. And so it 8 backs up, if you will, the passive system.
9 And it was decided by the designer. The 10 designer could have decided the shutdown system could 11 be the safety system and the RCCS, reactor cavity 12 system, could have been non-safety. But they made the 13 decision to go the other way.
14 So there were three ways to remove heat, 15 two, you know, one engineered and safety, one passive 16 and safety, and one engineered but non-safety. And 17 that would be consistent with how you described, you 18 know, single failure of a system.
19 MR. RECKLEY: Right, right.
20 MEMBER PETTI: Okay.
21 MEMBER KIRCHNER: But in your case, which 22 I know well, Dave, there were two safety grade systems 23 for the function. And as you said, they made the 24 decision.
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97 1 cavity, passive system versus the active shutdown 2 system that would get them down, you know, to low 3 enough temperatures for refueling. But there were two 4 safety related systems that could take the heat out of 5 a core.
6 MEMBER PETTI: Right. Yes, yes. But I 7 think that, the defense-in-depth requirement says no 8 single individual system. That's what the words say.
9 So this idea of having only one system would fail, 10 right --
11 MEMBER KIRCHNER: No, I --
12 MEMBER PETTI: -- from that requirement --
13 MEMBER KIRCHNER: -- Bill earlier 14 correctly that second or third system might not be 15 safety related.
16 MR. RECKLEY: Right. And you also, even 17 had to be careful in the example, because the pressure 18 boundary might have been safety related for another 19 purpose other than heat removal. So --
20 MEMBER MARCH-LEUBA: Yeah, on all the 21 defense-in-depth you're hanging your fruit on, your 22 hat on, it is so full of examples and such as and 23 maybe you can use margin and maybe you can use hand 24 waving. It's not clear that defense-in-depth says 25 anything honestly.
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98 1 If you want to take credit for defense-in-2 depth for your backup, your safety grade systems, you 3 have to tighten up the language and make sure that 4 defense-in-depth truly exists, because right now you 5 have so many qualifiers and examples and how you can 6 do this, how you could do that, that I don't have any 7 good feeling that it exists. Thank you for listening 8 to my complaints.
9 MR. RECKLEY: Okay. Back to 35, Marty.
10 MR. STUTZKE: Yeah, on the last row, I 11 would just point out that this notion of replacing the 12 single failure criteria with the reliability criterion 13 has been around about 18 years. And the Commission 14 approved it back in 2003.
15 Slide 36, please, a little history on 16 single failure criteria I thought that would provide 17 some perspective.
18 Back in 1965, the Atomic Energy Commission 19 convened a regulatory review panel to look at ways to 20 review policies and practices for licensing with an 21 eye towards expediting the licensing process.
22 And the panel came back. And one of the 23 recommendations was they felt there was an absence of 24 definitive requirements and criteria. And so to that 25 end, the Atomic Energy Commission proposed the general NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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99 1 design criteria in late 1965. They weren't finalized 2 until 1971.
3 But later on in '77, as the slide points 4 out, the Commission asked the staff to critique the 5 use of the single failure criteria. And the staff 6 said, yeah, it seems to be working, however, it's just 7 one of multiple fuels that are applied in system 8 design and analysis with the not comment, the single 9 failure criteria in and of itself is not sufficient.
10 They also pointed out the single failure 11 criteria was developed without testaments of 12 probability, some components or system failures.
13 Most importantly, they picked up on the 14 insights from WASH-1400, the original nuclear plant 15 PRA, and said things such as systems interactions, 16 what we would now call dependent failure analysis, 17 multiple human errors, tests and maintenance, all of 18 these things have an influence on reliability. And at 19 the time, they're not considered within the scope of 20 the single failure criteria, so we have to use 21 additional methods.
22 And one thing that I found very 23 interesting, almost prophetic, it says, gee, the use 24 of probabilistic methods such as the reactor safety 25 study, could be, areas could be increased and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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100 1 ultimately supplant the single failure criteria.
2 So, when you think about it, that was 3 some-40, well over 40 years ago. Okay.
4 MEMBER BLEY: I'd like to fill a little 5 bit more in on your history there, Marty. Back when 6 they did the reactor safety study, most folks here 7 weren't -- you know, they hired a bunch of guys from 8 Boeing to come over and bring folks and analysis with 9 them.
10 And those guys have (audio interference).
11 But then when we analyze these systems, you're going 12 to find that even more single point failures than we 13 would have ever guessed and reliability is much lower 14 than we expected.
15 That turned out not to be true. And it 16 turned out not to be true because of these single 17 failure criteria and the way the staff at that time, 18 you know, I've brought in some system analyses to talk 19 with the old generation of staff on what, dependent 20 failures.
21 They did track down some of these repeated 22 interrelated, interacting system failures and really 23 developed a deep questioning to look for single 24 failures. And that served very well.
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101 1 those failure in likely failures in systems. So it 2 has worked well. But it turns out that it missed some 3 of the more important things that were identified as 4 important in the study.
5 Anyway, I'm sorry, Marty. Go ahead.
6 MR. STUTZKE: No, you said it very well.
7 That was all that I was going to comment on this 8 slide. So slide 37, I think this one is yours, Bill.
9 MR. RECKLEY: Okay. One of the other 10 questions that came up during the subcommittee was 11 codes and standards and the phrase generally accepted 12 codes and standards.
13 So, since that time, we have released some 14 definitions. And one of the definitions we released 15 was of consensus codes and standards, which I won't 16 read here.
17 But basically it is our general 18 understanding. It's coming out of a standards 19 development organization and run through the normal, 20 you know, processes of ASME, ANS, ANSI, so forth.
21 So the, in terms of the discussion box 22 down below, we wanted to continue to encourage and 23 actually are required to encourage the use of 24 consensus codes and standards. So that's one of the 25 reasons we put the language in, to satisfy the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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102 1 National Technology Transfer and Advancement Act.
2 We do, however, recognize that there's a 3 lot of possible technologies in play and a lot of 4 different potential standards and somewhat of an 5 interest to also look at other standards approaches 6 like the International Standards Organization, or ISO, 7 and their standards in some areas for some components, 8 as well as the possibility of looking at other 9 international standards, if it happens to be a vendor 10 or a designer that's looking let's say to deploy in 11 Europe first, or some other area where another set of 12 standards other than, for example, ASME or IEEE might 13 be the ones generally used.
14 So, given the whole host of potential 15 standards, that was another reason we stuck with 16 wanting to encourage the use of consensus codes and 17 standards but not incorporate into the rules specific 18 codes and standards like the boiler and pressure 19 vessel code that we have for light water reactors.
20 We would look to, I think as we discussed 21 during the subcommittee meeting, look at guidance 22 documents, the submittal of proposals from either 23 SDOs, which we currently do, or the individual 24 designers or others and to try to pick that up in 25 guidance.
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103 1 And a recent example of that, for example, 2 is the Division 5 of ASME for high temperature 3 materials that we're looking to pick up in a Reg Guide 4 but not necessarily incorporate into 50.55(a).
5 And the last bullet there, one of the 6 reasons for that is that the incorporation of those 7 consensus codes and standards into the regulations has 8 raised other issues, including the need to do 9 rulemakings when they come up with new versions of the 10 codes. And that would be a little easier to handle in 11 guidance updates versus rulemakings.
12 So that was the slide on consensus codes 13 and standards and why you're not seeing ASME or IEEE 14 or ANS standards incorporated into Part 53, at least 15 where we are with the preliminary language.
16 So, if we go on into 38 --
17 MEMBER HALNON: Okay. Bill, just --
18 MR. RECKLEY: Yeah.
19 MEMBER HALNON: -- this is Greg. Just one 20 point on the consensus standards, in the guidance, you 21 know, I think it's a good idea, because it is poised 22 to get it into the rule and it takes a long time.
23 But do you foresee possibly taking any 24 kind of major exceptions to portions of the code? I 25 mean, that concerns me a little bit where the Reg NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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104 1 Guides come out and they'll have exceptions and/or, 2 you know, a differing thought process on a certain 3 endorsement. And you say we'll endorse, you know, 4 nine-tenths of it but not this last tenth of it. And 5 that could circumvent the use of it, the way it was 6 intended to be used.
7 MR. RECKLEY: It can. We always reserve 8 the right to do what you're saying, to put in 9 exceptions or clarifications.
10 Generally, we're able to avoid that in 11 many cases, and keeping in mind that often NRC people 12 are on the consensus code and standards, so we can at 13 least recognize what's coming and sometimes even 14 influence what's in the standard itself.
15 But, so hopefully -- I agree with you.
16 Hopefully, we could avoid that. And we traditionally 17 have avoided it in large part. But we do need to 18 maintain the ability in the development of a Reg Guide 19 to take exceptions to anything in a consensus code and 20 standard.
21 MEMBER HALNON: Okay. I guess it can be 22 important to point out that all that would go through 23 public comment, in addition to probably ACRS review as 24 well for --
25 MR. RECKLEY: Right, right. Yeah.
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105 1 MEMBER HALNON: Yeah. So, okay. Thanks.
2 MR. RECKLEY: Yes. All right. Okay. 38, 3 on the PRA, again, the requirement is being maintained 4 in our second iteration of the language to require PRA 5 to be done. And the use of the PRA as is highlighted 6 there is to at least support the assessment against 7 the second tier safety criteria of meeting the QHOs.
8 And with that, I'll turn it over. I think 9 the next slide --
10 MEMBER BLEY: Bill, I wanted to go back to 11 the point Greg made. I wanted to support the staff in 12 this area, because I have not seen a case that I was 13 involved in where the NRC was considering adopting a 14 consensus code or standard in which the NRC didn't 15 have one or more people on the committee that was 16 developing the standard.
17 So they were very knowledgeable about how 18 it was developed and what the intent was. So I don't 19 think there's much chance that you, you know, lose the 20 intent.
21 Usually, the clarifications and exceptions 22 are cases where the standard wasn't strict enough for 23 what NRC thought was the appropriate --
24 MR. RECKLEY: Okay. Thank you, Dennis.
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106 1 committee you're not representing the NRC. I mean, 2 there's the practical thing that you work for the NRC 3 and you're volunteering to be on a standard. But you 4 don't represent the NRC. I think most people know 5 that. I just think it's worth --
6 MEMBER BLEY: That's a good point. So, 7 when you come back to the NRC, you can represent --
8 MR. RECKLEY: Yes. You don't forget what 9 you just sat through. That's exactly right. That's 10 the point.
11 So, with this, I think one of the things 12 that came out of this subcommittee was a need for a 13 bit more discussion on PRA. So I'm going to hand it 14 back over to Marty for slide 39.
15 MR. STUTZKE: Okay. This slide looks at 16 past and present uses of PRA. These are listed in 17 Standard Review Plan, Chapter 19, which typically 18 applies to LWRs. But in general, these uses would 19 also apply to non-LWRs.
20 So the first one, it's about identifying 21 severe accident vulnerabilities. That one comes from 22 the advanced reactor policy statement, which in turn 23 references the severe accident policy statement.
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107 1 bullet pointed out our second tier criteria would 2 embed the QHOs directly into the rule language. This 3 one comes, again, from the Commission's advanced 4 reactor policy statement.
5 A third one here is use of the PRA to 6 support environmental reviews, specifically the 7 evaluation of SAMDAs, severe accident mitigation 8 design alternatives.
9 Now, to be clear, Part 51 does not require 10 the use of a PRA, but this is the way that it's been 11 done in the past. And I refer you to Regulatory Guide 12 4.2 in general on the preparation of environmental 13 reports and this new interim staff guidance 29, which 14 talks about environmental reviews and SAMDAs with 15 respect to micro-reactors.
16 But we point out in order to implement the 17 methods in these things you require a full level 3 18 PRA. Because it's a consequence, the idea is to 19 compute consequences of accidents, monetize them, and 20 then compare to the cost of implementing the 21 corrective action. So, in that respect, it's similar 22 to a --
23 (Off mic comments.)
24 MR. STUTZKE: I would also point out that 25 all the plants that have been certified designs and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 1 combined licenses all have level 3 PRAs in order to 2 meet this environmental review and the use of SAMDAs.
3 So, in the fourth bullet, if you're going 4 to implement the LMP guidance in NEI 18-04 you're 5 using the PRA to select and classify SSCs, and inform 6 defense-in-depth evaluations.
7 Let's go to slide 40.
8 (Off mic comments.)
9 MR. STUTZKE: Request people to mute their 10 microphones, please.
11 Okay. On slide 40, for applications that 12 are not based on the LMP, a PRA could be used to 13 support the Ritnis (phonetic), the identification of 14 systems incorporated within the program, et cetera.
15 The results and insights to the PRA are used to 16 support ITAACs, tech specs, COL action items, and 17 things like this.
18 Of course, the PRA may be used also to 19 support other concurrent voluntary risk informed 20 applications that may be included within a license 21 application, for example, risk informed in-service 22 inspection, risk informed tech specs. All of these 23 things could be in there.
24 And lastly, the staff uses the results of 25 the PRA to inform the scope of the review. This was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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109 1 an action that came from former Chairman Jaczko and 2 Commissioner Apostolakis. And it's known as the 3 enhanced safety focus review approach that's explained 4 better in, again, in the SRP like this.
5 But the idea is to focus the staff's 6 review on what's important and do a smaller amount of 7 review for things that the PRA says are not important.
8 Last and not least, the results of the PRA 9 are used to support reactor oversight programs.
10 So, continuing with slide 41 --
11 MEMBER KIRCHNER: Could you go back to 39, 12 Marty? This is Walt Kirchner.
13 MR. STUTZKE: Yes, of course.
14 MEMBER KIRCHNER: You know, I just -- this 15 is just an observation from a non-practitioner of PRA 16 but one who appreciates it.
17 I think the most important use of a PRA is 18 to gain insights and inform the design. And that 19 rarely ever gets listed.
20 It seems like the PRA is being used more 21 to determine regulatory compliance, to exclude things 22 from being on a D-RAP list, to, et cetera, et cetera, 23 which are all I think useful and important things.
24 But the most fundamental thing in my mind for the PRA 25 is to use the insights you gain to improve the design.
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110 1 And I just make that as a comment, that I 2 would hope that, to the extent that the rule is 3 requiring the PRA and such, that that is embedded in 4 the, one of the purposes, rather than I think there's 5 a tendency to, from my observing things on the 6 committee over the last five years, to focus on the 7 numbers and then use those numbers to exclude things 8 from regulatory treatment or, et cetera, et cetera.
9 And that's all justifiable. There's 10 economic reasons behind that.
11 But, again, I feel the most valuable part 12 of a PRA will be at the design phase to help inform 13 the design. And that rarely is cited.
14 MEMBER BLEY: I should, I really disagree 15 with your last statement. I agree with, 16 indefensible. But at least 4 of the design certs 17 we've done in the last 15 years made heavy use of 18 their PRA in the design process. In fact, it's what 19 led them to the new designs they proposed and got in.
20 (Simultaneous speaking.)
21 MEMBER KIRCHNER: I stand corrected.
22 MR. STUTZKE: Absolutely. So we expect in 23 the, is that the Commission argues in its advanced 24 reactor policy statement that it's clear the 25 Commission's intent was for designers to use the PRA NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 1 as part --
2 MEMBER KIRCHNER: That's where I agree.
3 That's I guess what I was getting at. And I stand 4 corrected by Dennis.
5 But, yeah, when I think of the 2008 6 advanced reactor policy statement, it's just some of 7 those concepts, if they were -- I think many of them, 8 of those are embedded already in your language that 9 you've been developing.
10 But that one just doesn't stand out to me.
11 Maybe I'm missing it somewhere or maybe, as Dennis 12 says, it's just done and that's it. But --
13 MR. STUTZKE: We'll take it under 14 advisement.
15 Okay. Another thing that I wanted to 16 discuss here is, in the letter the ACRS wrote on our 17 Part 53 white paper back in September of last year, 18 2020, you all used the phrase it's important to be, to 19 search for events without preconceived expectations.
20 And I know the topic had come up before 21 about how do you know that you're complete. So I 22 wanted to provide you with some language or some 23 thoughts that come out of the non-light water reactor 24 PRA standard, the various requirements on how the 25 initiating events are selected and how one confirms or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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112 1 attempts to show that they're complete like this.
2 So there are requirements to identify 3 initiating events, which are defined as challenges to 4 plant operations, and mitigate those challenges such 5 that you prevent a radioactive release.
6 That's put in there to account for things 7 like you may have, for example, a loss of feedwater, 8 followed by failure to scram. So the feedwater event 9 would become the initiating event, and the scram 10 failure and ATLAS sequence is treated elsewhere in the 11 PRA.
12 The second requirements are using a 13 structured systematic process. And it specifically 14 lists things like master logic diagrams, heat balance 15 fault trees, a process hazard analysis, failure modes 16 and effects analysis.
17 The process hazards analysis, the PHA, has 18 been a subject of study by the Electric Power and 19 Research Institute and its contractor, Vanderbilt 20 University. And they have actually applied it to the 21 old molten salt reactor experiment design to use it as 22 kind of -- I think of it as the prelude to the PRA, so 23 a very good process. They've issued reports on this.
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113 1 that the PRA standard specifies what to do. But it 2 doesn't tell you how to do it. Rather, it makes 3 references to other techniques that could be used like 4 this.
5 So, down to the third bullet, analyzing 6 operating procedures and practices to see where humans 7 could become involved and inadvertently trip the plant 8 off the line.
9 The fourth bullet is still in the 10 standard, review existing list of known initiators 11 specific to type. Obviously, that bullet by itself is 12 not sufficient.
13 One could come and say, take a list of LWR 14 initiators and say, gee, I'm designing a molten salt 15 reactor, so that one doesn't apply, that one doesn't 16 apply. And you don't -- ultimately you end up with 17 very few initiators.
18 So it's the totality of all these 19 requirements on this slide and the next one is what 20 provides the confidence.
21 That being said, conferring or referring 22 to known list of initiating events is an appropriate 23 way to do it.
24 MEMBER BLEY: Marty?
25 MR. STUTZKE: Yes.
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114 1 MEMBER BLEY: Comments. This is very 2 good. To my knowledge, the non-light water reactor 3 care is standard.
4 Has it been adopted as yet?
5 MR. STUTZKE: We are in the process of 6 endorsing that in a regulatory guide that will look 7 very similar to Reg Guide 1.200.
8 MEMBER BLEY: Yeah. And we'll see that 9 sometime. But, it isn't there yet.
10 MR. STUTZKE: Right.
11 MEMBER BLEY: The take we were trying to 12 make is that, of course, you should look at existing 13 lists, your last four.
14 But, really, that should be the last thing 15 you do. That should be a check on was there anything 16 in your other processes that you found out earlier?
17 If you're stuck with that list, it gets 18 harder and harder to really dig in for these other 19 approaches to try to make sure you're complete.
20 Anyway, but personally, I agree it belongs 21 on the list. But, I think it belongs at the end after 22 you've done the creative work of working hard for --
23 and using the things that might be hiding in your 24 design.
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115 1 out. But, this is the first time to my knowledge.
2 It's not the first time people have done this, it's 3 the first time to my knowledge it's been on any kind 4 of documents.
5 MR. STUTZKE: Yeah. That standard we 6 intend to go to start the Reg Guide publication 7 process towards the end of June or early July.
8 I know we have a meeting set up with one 9 of the subcommittees of ACRS to talk about it. But, 10 we hope to issue that standard by December, or endorse 11 that standard by December of this year.
12 MEMBER BLEY: That's good news.
13 MR. STUTZKE: On your last comment about 14 you were referring to the known initiators. But 15 personally, I've always looked at that.
16 That's like when you do, you know, a 17 calculus problem in school, and you know the answer is 18 in the back of the book.
19 So, you do all of the creative work up 20 front, and then you look in the back of the book and 21 see if you got it right. Something extra that maybe 22 you should have thought about.
23 MEMBER BLEY: Well, it certainly is. But, 24 I'll give you one, an anecdote, and this comes from a 25 lot of research in the area of expert elicitation.
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116 1 And there's been a number of studies where 2 they looked at the problem of anchoring, which is the 3 problem if you start with the last bullet. And with 4 a couple of the studies, they got people together and 5 they said, you know, just to help you out, we're going 6 to make up a first starting point.
7 And then you think about what could make 8 it more like or less likely. You know, you work from 9 that point.
10 But, now it's just made up. And if you 11 start with that, it's amazing how close you stay to it 12 by the time you've done the process.
13 You really don't want to bias yourself to 14 some anchor point where you've been searching broadly.
15 MR. STUTZKE: All right.
16 MEMBER BIER: Dennis, this is Vicki. If 17 I can just expand on that.
18 There was one study that specifically did 19 this for fault trees. Where they had like auto 20 mechanics or something.
21 And some of them looked at a complete 22 fault tree for why a car might fail to start. And 23 some of them looked at a fault tree when the cap 24 causes were missing.
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117 1 completed. So.
2 MEMBER BLEY: That's their problem. Okay, 3 thanks.
4 MR. STUTZKE: Okay. Let's go to slide 42, 5 please. Okay. So, in addition to the previous slide, 6 the list continues.
7 It says, you know, don't forget about the 8 external hazards. Including combinations of hazards 9 like seismically induced fires.
10 Looking at operating experience from 11 similar plants if it's available. Basically a 12 systematic evaluation down to the subsystem of the 13 train level.
14 Including all of the supporting systems.
15 So, you really understand the dependencies that the 16 gear system has with other systems, and things like 17 that.
18 Including initiating events that may have 19 involved multiple failures if they arise from a common 20 cause. That picks up things like earthquakes, these 21 big ones like this.
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118 1 of radioactive material.
2 The non-LWR PRA standard would consider 3 that you could have multiple reactors onsite, plus 4 non-reactor radiological sources.
5 So, spent fuel or off cast systems, things 6 like that. And they would all be included in the 7 scope of the PRA.
8 So, okay. Moving to slide 43. Another 9 question that we commonly have to address is, what 10 about the lack of operating experience?
11 So what I've tried to list here in the 12 lefthand column here, are all the, for lack of a 13 better word, the numbers that go into the PRA 14 calculation. The initiator frequencies, the component 15 failure rates, and so forth, is listed here.
16 And thinking about it, a great many of 17 them can be estimated using existing nuclear or non-18 nuclear information. We point out that a great deal 19 of the zeta that went into the original WASH-1400 20 study was from non-nuclear sources like this.
21 And they can be formally combined using 22 Razian statistical methods, which allow you to mix 23 limited sets of operating experience with subjective 24 judgments.
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119 1 expert elicitation on this. I would point out, these 2 are currently done for large light water reactor PRAs 3 as well.
4 That's where we get numbers like the 5 frequency of large break LOCAs, is through an 6 elicitation process.
7 Common cause failures, we have good 8 models, such as the Alpha Factor model that's been 9 used like this. We have very good generic information 10 that's been developed over a number of years.
11 One of the things that I would point out, 12 what is interesting about the generic common cause 13 failure data is it's stability in the sense that 14 numbers don't change among systems too much, or 15 components too awfully much.
16 So, it's reasonably robust. Yes, Dr.
17 Bier?
18 MEMBER BIER: (No response) 19 MR. STUTZKE: Is there a question from 20 Vicki?
21 MEMBER BIER: Yes. I had to unmute.
22 Sorry. I just wanted to chime in again on the topic 23 of expert elicitation.
24 And, I think this is again for background.
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120 1 incorporated in this document.
2 But, over time there is a lot of evidence 3 that not all experts are equally good at putting what 4 they know into probabilistic terms.
5 So, at some future time, the Committee or 6 the Agency may want to look into updating the guidance 7 on expert elicitation. But, I don't think that needs 8 to be part of this process today. Thanks.
9 MR. STUTZKE: Yes, thank you. It reminds 10 me of an interaction I had once with former Commission 11 Apostolakis.
12 He told me, when he estimated numbers for 13 use in a PRA, he was providing his expert opinion. On 14 the other hand, if I estimated the same number, I was 15 just guessing.
16 So, you're right. Different experts have 17 different qualifications. And I would agree, we need 18 to revisit our guidance on how to conduct expert 19 elicitation.
20 Jumping down to the bottom of the list 21 there, human error probabilities, hazard frequencies, 22 external hazard fragilities, none of those require 23 design specific operating experience.
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121 1 methods for performing those sorts of analyses.
2 Any other questions on this slide? Vicki?
3 MEMBER BIER: (No response) 4 MR. STUTZKE: Okay. Well, let's go onto 5 slide 44 then. But, I would also emphasize the PRA 6 provides a framework for assessing the uncertainties 7 normally lumped into the parametric uncertainties, the 8 modeling uncertainties, and the completeness 9 uncertainties, like this.
10 We would certainly expect that people 11 don't just estimate the uncertainty and all the 12 parameters. Do a Monte Carlo propagation up to the 13 final risk metrics and call it a day.
14 They're actually obliged to understand 15 what factors, which basic events, human error, et 16 cetera, et cetera, are driving the uncertainty in the 17 overall results. So, kind of a decomposition.
18 But, that's the process that helps you put 19 the uncertainty into perspective. As you can see, 20 things that might be uncertain or questionable, let's 21 say, because of a lack of operating experience.
22 For example, the turbine trip rate, or an 23 uncomplicated scram rate. Things that we would 24 normally estimate using a lot of statistical data, but 25 we would lack, because the plan hasn't been built.
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122 1 Those sorts of events don't tend to be 2 risk significant. And therefore, they won't overall 3 have too great an impact on the final risk calculation 4 or the uncertainty in that risk calculation.
5 Okay. Comments on that?
6 MEMBER DIMITRIJEVIC: Yes, Monty, I have.
7 Okay. I really appreciate, you know, how you well 8 summarize the usefulness of the PRA. That was really 9 good.
10 But, the new list of all the PRA 11 applications, the only one with actually level three 12 results were necessary for any of the report, and some 13 -- and most of the design certification which have 14 been submitted now, are not required to have a level 15 three PRA.
16 So, now when we came to the -- to this 17 slide of uncertainties, you also nicely summarize the 18 positivity. Because one of the main issues with using 19 PRA is associated with uncertainties.
20 So, we don't even have a good way to 21 address every item that was in the PRA. Which 22 completely addressed modeling uncertainties or 23 completeness uncertainties.
24 And those uncertainties are still open.
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123 1 you agree with me the uncertainties multiply like, you 2 know, half the times responded with level one PRA.
3 And doing the level three, especially when 4 you don't have a location, increases uncertainties 5 associated with the results significantly.
6 And this is my main objection. Is why do 7 we want to introduce these QHOs when we actually 8 really, you know, the drop are uncertain?
9 MR. STUTZKE: Well, I would respond, you 10 know, we've done -- the staff has done extensive 11 looking and the state of the art reactor consequence 12 analysis, the SOARCA Project.
13 Which included a full propagation of 14 uncertainty all the way through the MELCOR and the MAX 15 codes. And the uncertainties were perhaps not as big 16 as one would expect.
17 All right. The other thing is that the 18 Commission's safety role policy statement, while it 19 was being developed, considered how to decide whether 20 somebody had met the goal.
21 And after a lot of discussion, they 22 concluded the best way was to compare the mean of the 23 uncertainty distribution to the care chart as well as, 24 you know, then later consider the uncertainties as 25 I've described here on slide 44.
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124 1 So, from that perspective, you know, the 2 issue was debated a while back. And reasonably result 3 --
4 (Off record sound interruption) 5 MEMBER BLEY: Okay. If the members on the 6 public on the line can mute your phones.
7 (Off record sound interruption) 8 MEMBER BLEY: Quynh, if you can help us 9 out.
10 MR. NGUYEN: The members on the public 11 line with a radio or music, could you please turn it 12 off?
13 MEMBER BLEY: Thank you. Marty, can you 14 hear me now?
15 MR. STUTZKE: (No response) 16 MEMBER BLEY: Or did Marty drop off?
17 MR. STUTZKE: I am here Dennis.
18 MEMBER BLEY: Oh good. Okay. I want to 19 take you back to your slide 41. You've got me 20 curious, and I started digging around.
21 I really like this. But, I'm remembering 22 back some time in the last year or so, maybe it was 23 two, because I think we were in person back then in 24 Rockville.
25 But, we had a meeting with the staff and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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125 1 with representatives of the committee developing this 2 non-light water reactor PRA standard. And we 3 challenged them that there was no guidance here like 4 in other places, on how to structure this search.
5 And I think we said a structured 6 systematic process. And what date is the version on 7 the standard that you have?
8 Because I've got the one we reviewed back 9 then, with a 2020 date on it. And I can't find any of 10 this in there.
11 And the representatives from there said 12 that if you think about it, that they couldn't speak 13 for the committee. So, maybe it's just recently been 14 developed.
15 But, the bottom line is, I'm glad to see 16 it's going to be here. But, I don't think it was 17 there a year and a half ago or so, whenever we had 18 that meeting.
19 But, I'm pleased it's here, so.
20 MR. STUTZKE: Yeah. I don't know about 21 that. The final version of the standard that was 22 issued in February 2021.
23 MEMBER BLEY: Well, that's really great.
24 And they sent me a bunch of papers that are right in 25 line with this, as to how they've used it in other NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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126 1 analyses.
2 And they put it in here. So, I'm glad to 3 see. I guess it's under initiating events?
4 MR. STUTZKE: Right.
5 MEMBER BLEY: We'll have to get their 6 version.
7 MR. STUTZKE: Yes.
8 MEMBER BLEY: But, use this -- is this 9 their interpretation?
10 MR. STUTZKE: Yes. This is a compilation 11 of the various supporting requirements.
12 MEMBER BLEY: Okay. Derek, we'd like to 13 get that. We're out of date. Thank you.
14 MR. WIDMAYER: I heard you.
15 MR. STUTZKE: Well, with that, I'll turn 16 it back over to Bill. I do have some backup slides on 17 PRA and risk metrics if you would like me to discuss 18 any of them.
19 The origin of the QHOs, the risk 20 surrogates, things like that.
21 MEMBER BLEY: All right. I think some of 22 us would be interested. But, unless other members 23 really want to see it, I don't think we'll go to that 24 today. I don't think it's terribly relevant.
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127 1 presentations on some of the materials that you guys 2 have talked about today. And I just wondered, is all 3 of that something you could bring and present?
4 Or is some of this available in some kind 5 of self study modules at the Commission?
6 MR. RECKLEY: It is. It's publically 7 available. Most of it, I don't know how it could be 8 -- I don't know if it would be effective in self study 9 mode.
10 MEMBER BLEY: Okay.
11 MR. RECKLEY: We can -- we can provide it 12 to you, and you can maybe help us assess --
13 MEMBER BLEY: If you can get that to 14 Derek, I'll take a look. And then we can upload it to 15 the rest of the members and see if anybody wants this 16 in study mode.
17 MR. RECKLEY: Okay. We will provide that.
18 MEMBER BLEY: Thank you.
19 MEMBER KIRCHNER: Dennis, this is Walt.
20 Let me ask you a question.
21 Based on what Marty's just presented, does 22 this address what you've often stated, starting with 23 a blank sheet of paper, and doing a completed search 24 -- well, not a complete search, but a well-informed 25 search or initiating events, and defining design basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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128 1 events?
2 MEMBER BLEY: Yes, it does. Although on 3 the slide, it wasn't quite expressed that way. And I 4 don't know what it says in the standard. But, we 5 better look at that sometime soon.
6 But, if you remember, we had that meeting 7 on the standard. They were going to come back at some 8 point.
9 And we talked about this issue. And the 10 representative that was there, I used to work with, 11 sent us about 20 papers dealing with these issues.
12 And they're the same kinds of things I was 13 putting together in that White Paper we covered. So, 14 maybe it will save me some effort and not going to 15 read it.
16 But yes, it's supposedly reentered if it's 17 not the same.
18 MR. RECKLEY: Okay. We have a few more 19 slides to finish up. But, I guess before getting into 20 that, I'll apologize on the single failure 21 discussions.
22 It's the danger of trying to do by 23 examples. But, I'll also look to see if there's some 24 clarification.
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129 1 involved on the periphery, but the design review guide 2 that the ACRS did look at, it talks about INC, and 3 whether INC is developed under an LNP approach, or the 4 more traditional single failure.
5 I think both of those avenues are 6 addressed within that design review guide again, that 7 the ACRS has looked at.
8 So, I'll gather up some examples of that 9 as well. And then maybe even some past examples.
10 MEMBER BLEY: Was that -- was that the 11 read list of Chapter 70 SRP? Or --
12 MR. RECKLEY: Yes.
13 MEMBER BLEY: Okay.
14 MR. RECKLEY: Yes, that. So, finishing up 15 on the last few slides under design and analysis. We 16 did revise the guidance from the first time that ACRS 17 looked at it.
18 This is consistent with what we brought to 19 the subcommittee. And for the purposes of design and 20 licensing basis event selection, safety class and 21 SSCs, that other engineering approaches could be used 22 for that.
23 That was -- this change was a result, or 24 resulted from public stakeholders who wanted to make 25 sure we weren't foreclosing on other generally NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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130 1 accepted ways, be they reflected in other guidance 2 documents, IAEA approaches, and so forth.
3 So, we did change this. I think it really 4 doesn't change much in the way of the overall 5 requirements or approach. Go to slide 46.
6 We did go through some iterations on some 7 of the specific sections. If you want to go to 47, 8 that's one example.
9 And again, we brought this before the 10 subcommittee. I'm not sure it was -- felt at that 11 time it warranted a lot of discussion.
12 But, we did expand and tried to clarify 13 that the -- that an applicant would need to look at 14 the whole range of licensing basis events from AAOs 15 down to very unlikely events.
16 Go onto 48. This is the DBA. And there 17 was some discussion of this at the subcommittee 18 meeting.
19 We did add a specific sentence that said 20 for the DBAs they needed to be analyzed from 21 initiation to a safe stable end state. And again, as 22 we've talked several times, assuming only safety 23 related SSCs and safety-related human actions would be 24 credited in that assessment of the DBA.
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131 1 the subcommittee meeting as to whether this kind of 2 scratched that itch that I think was identified in the 3 first iteration?
4 (No response) 5 MR. RECKLEY: Okay. Seeing no hands, 49, 6 slide 49. Just kind of wanted to emphasize that we 7 did maintain a fairly traditional safety 8 classification scheme of having safety related, non-9 safety related but safety significant, which for those 10 more familiar with LNP, those would be non-safety 11 related with special treatment, and non-safety 12 significant SSCs.
13 And you can draw parallels between that 14 and some of the other approaches like regulatory 15 treatment of non-safety systems. Or the primary, the 16 three prim -- three of the four risk categories in 17 50.69.
18 And even to some degree, some 19 similarities, when you start to look at IAEA, specific 20 safety requirements and the introduction of design 21 extension conditions.
22 So, if there's no questions on safety 23 classification, we can go to slide 50.
24 MEMBER KIRCHNER: Is there a --
25 MR. RECKLEY: Go ahead.
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132 1 MEMBER KIRCHNER: Bill, this is Walt 2 again. Is there any need to make it clear somehow 3 that like if you're invoking the LNP approach, the PRA 4 and all the rest that, I -- you said it, and I can't 5 remember it.
6 They don't -- the middle category, they 7 call it non-safety related but risk significant? Or 8 -- I can't remember.
9 MR. RECKLEY: Yeah. Under LNP or in NEI 10 1804, it's called non-safety related with special 11 treatment.
12 MEMBER KIRCHNER: Special treatment.
13 Sorry, I misspoke.
14 MR. RECKLEY: And it really is equivalent 15 to what we're calling non-safety related but safety 16 significant.
17 In that what we -- what we'll ultimately 18 say is needed for any SSC that's designated as non-19 safety related but safety significant, is the 20 definition of what is needed in terms of special 21 treatment.
22 Be it hardware requirements like the 23 environment it needs to withstand, be it relia --
24 almost certainly a reliability assurance program and 25 measure to carry through in operations.
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133 1 And maybe even programmatic requirements 2 in terms of inspections, procedures to operate the 3 equipment. Whatever special treatment is needed in 4 order to ensure that that SSC --
5 MEMBER KIRCHNER: Right.
6 MR. RECKLEY: Would have the capabilities, 7 the reliabilities, the availabilities that are assumed 8 in the assessments.
9 So --
10 MEMBER KIRCHNER: Okay. And so, the --
11 obviously then, these will make your definition table 12 some place.
13 MR. RECKLEY: Yes.
14 MEMBER KIRCHNER: And then we won't have 15 to deal with the many other terms. At least in terms 16 of 53, at least.
17 And we would -- we would just have these 18 three. We wouldn't have the two by two box. We 19 wouldn't have other -- other terminology then in 53.
20 We would be self-consistent.
21 MR. RECKLEY: Yes. That's the goal.
22 MEMBER KIRCHNER: Good. Good.
23 MR. RECKLEY: Yeah. And there were 24 certain terms that we avoided on purpose, just to not 25 carry forth the confusion for another 50 years.
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134 1 So, with that, I think we can -- if 2 there's no more questions on that. Analytical margins 3 for operating flexibilities, we've talked about this 4 a fair amount.
5 And we've really not changed the language 6 very much. Or at all really. This is the provision, 7 this is the section that would define how the analysis 8 needs to be carried through and maintained to support 9 something like the calculation that you could have a 10 smaller emergency planning zone, or you could justify 11 an alternative to the population density criteria in 12 the siting reg guide.
13 Any other operational flexibilities that 14 we're going to start to get into in Subpart F, this 15 provision is allowing the margins to be traded off.
16 And then establishing the requirements to 17 make sure that all the assumptions and analysis that 18 went into justify trading off the margins, are 19 maintained over the life of the plan.
20 So, we can go onto 51, I think. There 21 were really no changes or much of a discussion with 22 external stakeholders or with the ACRS subcommittee on 23 the need to have quality assurance for the design 24 process, and the need to set up interfaces between the 25 design process and things like construction, fairly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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135 1 obvious, operations, and so forth.
2 So, I don't -- I won't spend much more 3 time on that. Slide 52 goes to non-radiological 4 hazards. And we talked about this with the 5 subcommittee.
6 We're just -- we are still looking at this 7 and looking at other examples like fuel cycle 8 facilities, to see how we should bring in the non-9 radiological hazards into Part 53.
10 We'll acknowledge that it warrants, I 11 don't know, reviewing that topic. And we're currently 12 doing that. And we will come back to the ACRS on our 13 resolution of that.
14 And with that, I think that's the last 15 slide. Yes.
16 MEMBER BLEY: Okay. Bill?
17 MR. STUTZKE: Yes, Dennis?
18 MEMBER BLEY: I might go back to Marty if 19 he's still with us. Marty, could you pull up your 20 slide 64 and then 65?
21 MR. STUTZKE: Yes.
22 MEMBER BLEY: My first question is just 23 personal curiosity on 64. The surrogates here, the 24 QHOs, my memory is that this stuff was put together by 25 Trevor Bott (phonetic) and maybe John Lanoff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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136 1 (phonetic), but both may have been by going to 2 existing PRAs and kind of summarizing this.
3 Do you remember if that's how that came 4 about?
5 MR. STUTZKE: Yes. That is.
6 MEMBER BLEY: And then --
7 (Simultaneous speaking) 8 MR. STUTZKE: And when a --
9 MEMBER BLEY: Go ahead.
10 MR. STUTZKE: Yeah, that's how it was 11 done. I mean, they actually described it pretty well 12 in the Appendix D of the technology neutral framework, 13 NUREG-1860.
14 MEMBER BLEY: That's why I remember it.
15 Thank you.
16 MR. STUTZKE: Yes.
17 MEMBER BLEY: Go onto 65. There's some 18 real key stuff in this slide. Maybe you can talk us 19 through this one.
20 And you know, we've -- we thought some 21 about some difficulties. Now Bill's assured us, and 22 we've found they're not too difficult directly of 23 having the QHOs.
24 But, there are arguments about why it 25 might be better if you use other integral risk NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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137 1 measures to test the overall risk for a new plant.
2 And that you've cited something up here 3 that was also in Appendix in 1860. But, I think in 4 the body of 1860 they went with the QHOs as an 5 integral risk measure.
6 But in the Appendix they used various 7 other approaches. If you could talk us through that, 8 I think that would be helpful.
9 MR. STUTZKE: Yeah. It's an historic 10 issue as you said, Dennis. They ended up using the 11 QHOs in the main body.
12 And there is this extended discussion 13 about the use of complementary accumulative 14 distribution functions, CCDFs, in there. I've cited 15 the main ACRS letter where they debated, you know, the 16 members at that time about the pros and cons of using 17 the method.
18 The staff deferred action on it. One, 19 because the project was coming to an end. And they 20 wanted to get 1860 published.
21 The more technical reason is they were 22 worried about anchoring the CCDF to the QHOs. So, the 23 area under the CCDF is the main risk, right. The 24 expected value of risk.
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138 1 is then, how do you draw the line? You know, the 2 limit line that the CCDF would represent.
3 And they never got to that point during 4 the development of NUREG-1860. But rather, it was 5 intended to be deferred until they could pilot 1860 6 either on the, I think the pebble bed design.
7 MEMBER BLEY: Yeah. It was the pebble 8 bed. And they backed out.
9 MR. STUTZKE: Yeah. They backed out and 10 it never got done. It was never picked up.
11 MEMBER BLEY: So, have you given any 12 thought to relative merits of sticking with the QHOs?
13 Or using something like a CCDF limit curve?
14 MR. STUTZKE: Yeah. I thought about it 15 occasionally. About how would I come up with the, you 16 know, the shape of the CCDF curve. And make certain 17 it goes through appropriate anchor points and things 18 like that.
19 But, that's about as far as I've gotten.
20 I'm not ready to say, you know, one way is better than 21 the other.
22 MEMBER BLEY: Of course Rich Denning and 23 Vinod Mubayi, and Vinod was probably the primary 24 author of that Appendix, wrote a letter to the staff 25 on this very issue.
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139 1 MR. STUTZKE: Yes.
2 MEMBER BLEY: I saw a copy of the letter.
3 Is there a response to the letter? Yes, there is.
4 And it's in -- you got it there. That's the end on 5 that one.
6 MR. STUTZKE: Yes.
7 MEMBER BLEY: Could you tell us how the 8 staff responded?
9 MR. STUTZKE: Bill, you want to jump in 10 since I didn't write this response?
11 MR. RECKLEY: Our response was basically 12 to acknowledge that what Dr. Denning and -- had 13 proposed was a workable approach.
14 However, we also thought that the LNP, 15 looking at the individual events, and comparing it to 16 the frequency consequence, when combined with looking 17 at the cumulative risk through looking at the QHOs and 18 the other cumulative measure that LNP had provided, 19 that it was also an acceptable approach.
20 And since for the purpose of writing the 21 regulatory guide, we were being asked to endorse NEI's 22 1804, that -- that as Marty said, we weren't saying 23 one was necessarily preferable over another. Both 24 could work.
25 And so that was the response.
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140 1 MEMBER BLEY: Thank you. We may come back 2 to this some more later. Okay. So, I just wanted to 3 pick up those two.
4 I don't have anything else. Do any other 5 members have questions or comments you'd like to make 6 before we move to public comments?
7 MEMBER DIMITRIJEVIC: Well Dennis, I heard 8 two interesting things today. And I was wondering if, 9 are those two informations publically available?
10 And can we see some of that? Like for 11 example, there was a, I think Bill said there was a --
12 they've used the QHOs on some of them for Fukushima.
13 So, that would be something that would be 14 interesting to see. And also, Marty said that the 15 level three is out, which they've done.
16 I'm not sure I support that right at this 17 moment with the associate uncertainties, show the 18 uncertainties are not so high on level three results.
19 And I -- if that's a public available 20 book, I would love to see that.
21 MEMBER BLEY: All right. If the staff 22 could deliver any of that to Derek, that would be 23 helpful. My memory is that back when we reviewed the 24 Fukushima items on the failure to vent for all 25 reactors, that was part of the analysis, was that.
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141 1 But Bill, I'm not positive of that.
2 MR. RECKLEY: Yeah.
3 MEMBER BLEY: So were --
4 MEMBER RECKLEY: Yes, you did. And it was 5 subsequently published in a NUREG. And so we'll give 6 you that reference.
7 MEMBER BLEY: Okay. I think that would be 8 helpful. And that would be great for that.
9 MR. RECKLEY: Yes.
10 MEMBER BLEY: Okay. Thank you. And 11 Derek, you're on the line. I remember we got the 12 comments from Rich and Vinod.
13 Did we also get the staff response? Or 14 can you get that for us?
15 MR. WIDMAYER: I think I got it. But, 16 I'll check and make sure.
17 MEMBER BLEY: Okay. Thank you.
18 MR. WIDMAYER: Yeah.
19 MEMBER BLEY: Anyone else?
20 (No response) 21 MEMBER BLEY: At this time I'd like to 22 open the public line. Oh, no, I'm sorry. My day has 23 gone blank.
24 But, at this time I'm going to invite 25 Cyril Draffin from the USNIC to speak. They requested NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 time with the meeting to make some comments.
2 And if you are available, please begin.
3 MR. DRAFFIN: I am. Thank you very much.
4 I am Cyril Draffin, the Senior Fellow for Advanced 5 Nuclear at the U.S. Nuclear Industry Council. And 6 today's remarks augment the comments we provided at 7 the April 22 subcommittee meeting.
8 First, it may be premature for the ARC to 9 make a definitive comment. The NRC has stressed that 10 Part 53 permitting language will remain open to change 11 until all parts of Part 53 have been provided and 12 stakeholder comments have been received.
13 Therefore, it may have a negative impact 14 for ACRS to submit a definitive interim letter to 15 support the current Subpart B and C drafts of the 16 rule. Recognize that only a current portion of the 17 Part 53 language is available and the current language 18 is likely to change.
19 Second, the U.S. Nuclear Industry Council 20 does not agree with the second iteration of Subparts 21 B and C. There are many areas where the preliminary 22 language in Subparts B and C are increasing regulatory 23 burden over Parts 50 and 52.
24 And the NRC has basically made no major 25 changes to address the industry concerns about those NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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143 1 two Subparts.
2 Also, there is a -- the NRC staff had 3 promised that Subpart F would enable a significant 4 reduction in operational burden as compared to Parts 5 50 and 52.
6 And therefore, that justified the 7 increased burden in Subparts B and C to obtain those 8 operational duties, the operational burden.
9 But, now having seen Subpart F, it's not 10 clear what the benefits are. The preliminary language 11 seems to result in increased burden, doesn't -- still 12 limits flexibility, and doesn't really enhance the 13 safety.
14 So, we're hoping that the NRC will be 15 receptive to incorporating some of industry's 16 stakeholders' inputs in the coming months.
17 The only apparent benefit of Part 53 so 18 far, is that there's no need to seek exemptions to 19 large LWR specific requirements.
20 Then a few points that we've covered 21 before, which I think are still relevant, particularly 22 for people that weren't on the subcommittee meeting.
23 For the adequate protection standard, we 24 disagree with the second revision to the strategic 25 objectives that drop the formal reference to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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144 1 reasonable assurance of adequate protection standard.
2 We think that it, adequate protection to 3 public health and safety is important. And changing 4 the objectives primarily to justify the preliminary 5 language seems questionable.
6 For the tiers, we still think the tiered 7 category one and two are confusing, with opportunities 8 for unintended consequences. The second rendition of 9 the SOARCA objectives drops the language in the Atomic 10 Energy Act, and so the same 51 and 52 seem less 11 relevant.
12 And we might consider a simple tier unless 13 the operational language shows real benefits, and 14 particularly for all the criteria discussed earlier 15 that have to be met.
16 We continue to believe Part 53 should be 17 technology inclusive to allow both risk based and 18 deterministic methods. And that it should not be 19 limited to just applications using the PRA tool, 20 although it's a very valuable tool 21 And with this second iteration, it's still 22 too restrictive in requiring a PRA. As discussed 23 earlier, we think PRA should be applicable for a range 24 of licensing path and technologies.
25 And that risk insights are what are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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145 1 important for design, not the specific numerical 2 results. We don't believe that PRA should be elevated 3 to a compliance tool as part of the application, 4 especially for a construction permit.
5 It's not clear that any approach used by 6 Oklo or NuScale would comport with a prescriptive use 7 of the PRA as a compliance tool.
8 And if it is included, as it is not, in 9 Part 53 as a requirement, then exemptions will be 10 required for some of the technologies, which seems a 11 little inconsistent with the original goals and 12 objectives of Part 53.
13 Now, the timing for a phased or simplified 14 approach has merit. And I think that there's some 15 flexibility on how that's done. It merits further 16 discussion.
17 For ALARA, many stakeholders, as mentioned 18 this morning, believe that ALARA is an important 19 concept and certainly a good practice that we expect 20 to continue.
21 But, we do not believe ALARA should be 22 included in Part 53 formal regulation in part because 23 of the subjectivity and complexity of ALARA in the 24 design phase. New operation should be like protection 25 of plant workers and should not be included in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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146 1 safety criteria.
2 And for defense in depth, it's important 3 as a design philosophy and supporting an adequate 4 safety case. But, the defense in depth details should 5 be in guidance and not added to the regulations.
6 We believe that Part 53 can have 7 predictability as well as flexibility. We think it 8 really can have specific performance criteria that 9 must be demonstrated, and flexibility to allow them to 10 be made and not just relying upon LNP as the process.
11 And finally, we do support the consensus 12 codes and standards, which are being adopted by NRC.
13 So, those are some comments for you to consider as you 14 draft your interim letter.
15 Thank you for the opportunity.
16 MEMBER BLEY: Mr. Draffin, thank you very 17 much. You will be on the transcript. And you will 18 have access to that.
19 But, if you prefer to also send your 20 comments in writing, both they -- I shouldn't say it 21 won't be garbled, but once in a while, transcripts 22 don't read exactly like you thought you said them.
23 So, if you wanted to do it in writing too, 24 that's fine. Just give them to Derek.
25 MR. DRAFFIN: Well, thank you.
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147 1 MEMBER BLEY: So, Thomas or Makeeka, can 2 we get the public line open?
3 MR. DASHIELL: The public line is open for 4 comments.
5 MEMBER BLEY: Thank you very much. If 6 there's anyone who would like to make a comment, 7 please identify yourself and make your comments.
8 (No response) 9 MEMBER BLEY: All right. Okay, I think we 10 can close the public line at this point.
11 For the members, we had a very long 12 session last week of deliberations. And as a result 13 of that, I really thank everyone for all the ideas and 14 written suggestions and the discussion.
15 But, it helped a lot. As I began to 16 organize my notes from it, that session, the pieces 17 started to come together.
18 And I think they're -- they're still 19 pretty much holding out that there might be some areas 20 we'll have to dig into. As I drafted a letter, I 21 tried to include areas where I had a sense we had 22 agreement.
23 And for other areas, rather than 24 reconcile, I tried to integrate that, or put together 25 the areas of concern. I couldn't address every issue NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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148 1 that everyone raised. We don't need that in this 2 letter.
3 Added comments may be necessary or are 4 certainly welcomed. But that's kind of irrelevant.
5 You can write them anyway. But any area where one or 6 two people feel very strongly about, maybe it's 7 important.
8 At this time, I think we'll call this 9 meeting to an end. But, we'll move into a letter 10 writing session if that's okay, Matt.
11 But we will go off the record. But, I'll 12 let you do that since this is a full committee 13 meeting.
14 CHAIR SUNSERI: Thank you, Dennis. So, 15 could we take like a ten minute break though before we 16 go into reading the letter?
17 Would that be okay to everyone?
18 MEMBER BLEY: Yeah. I was going to 19 suggest that. And I think we'll try to finish in 15 20 or 20 minutes after.
21 I don't know, if we really get into 22 discussion, it could take a long time. But, I'd like 23 everybody to hear where it stands now and be able to 24 read it later.
25 I think Derek will verify this. I think NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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149 1 he's got it up on the website. But, I'm not quite 2 sure.
3 CHAIR SUNSERI: Okay. Very good. Then 4 we'll take a 15 minute break here. We'll recess until 5 6:15. Is it 6:15, is that right? Yeah.
6 Oh, it's 6:00. Okay. All right. We'll 7 recess until 6:15. And then we'll pick it up and read 8 through the letter and finish today at the conclusion 9 of that activity.
10 All right. So, we are recessed.
11 (Whereupon, the above-entitled matter went 12 off the record at 6:00 p.m.)
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Advisory Committee on Reactor Safeguards (ACRS) 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors May 5, 2021 1
Agenda
- Opening Remarks
- Overall Structure (Framework)
- Subpart B - Technology-Inclusive Safety Requirements
- Subpart C - Design and Analysis Requirements
- Discussion 2
Background
- Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439) signed into law in January 2019 requires the NRC to complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use for commercial advanced nuclear reactors no later than December 2027 o (1) ADVANCED NUCLEAR REACTORThe term advanced nuclear reactor means a nuclear fission or fusion reactor, including a prototype plant with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of this Act, 3
NRC Staff Plan to Develop Part 53 Subpart B Subpart C Subpart D Subpart E Subpart F Subpart G Project Life Cycle Design and Siting Construction Operation Retirement Requirements Definition Analysis
- Safety Objectives External Facility Safety
- Safety Criteria System Hazards Construction/ Program
- Safety Functions & Component Manufacturing Design Site Surveillance Characteristics Ensuring Maintenance Analysis Capabilities/
Requirements Environmental Reliabilities Configuration Considerations Control Safety Change Control Categorization Staffing &
& Special Environmental Human Factors Treatment Considerations Programs Security, EP Other Plant/Site (Design, Construction, Configuration Control) Clarify Controls Subpart A Analyses (Prevention, Mitigation, Compare to Criteria) and General Provisions Distinctions Plant Documents (Systems, Procedures, etc.) Between Subpart J Admin & Reporting LB Documents (Applications, SAR, TS, etc.) Subparts H & I 4
Part 53 Contents (A) General Provisions (including definitions)
(B) Safety Criteria (two tiers/categories, as low as reasonably achievable (ALARA), defense in depth (DiD)
(C) Design and Analysis (design criteria, role of probabilistic risk assessment (PRA))
(D) Siting (external hazards, population)
(E) Construction and Manufacturing (factory fueling)
(F) Operations (structures, systems and components (SSCs),
staffing, programs)
(G) Decommissioning (H) Licensing (siting, design, licenses)
(I) Maintaining Licensing Basis (J) Administrative and Reporting 5
Subpart B Technology-Inclusive Safety Requirements Preliminary Language 6
Rulemaking Plan (SECY-20-0032)
The staff plans to build upon ongoing activities such as those described in SECY-19-0117, Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, dated December 2, 2019 (ADAMS Accession No. ML18311A264), to develop the associated performance criteria.
The methodology described in SECY-19-0117, includes identifying the potential benefits provided by design features and programmatic controls in terms of the margins between estimated doses and the reference values in NRC regulations and the margins between estimated health effects and the NRCs safety goals. SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, dated September 28, 2018 (ADAMS Accession No. ML18115A157), and SECY-18-0103, Proposed Rule: Emergency Preparedness for Small Modular Reactors and Other New Technologies (RIN 3150-AJ68; NRC-2015-0225, dated October 12, 2018 (ADAMS Accession No. ML18134A076),
provide examples of how those margins are used within performance criteria for potential operational flexibilities.
7
First Principles See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities, and SECY-19-0117, Technology-Inclusive, Risk-Informed, and Performance-Based Methodology..
8
Subpart B - Safety Criteria
- Safety Objectives
- First Tier Safety Criteria (B) Safety Criteria What function(s)
- Immediate threat to public health (e.g., a barrier, cooling) and safety are needed to satisfy safety criteria
- Second Tier Safety Criteria
- Appropriate to address potential (B) Safety Functions risks to public health and safety What design features (e.g., a structure, system)
- Safety Functions are provided to fulfill the safety function(s)
- Licensing Basis Events (C) Design Features (LBEs) (and Human Actions) What design criteria (e.g., leak rate, cooling
- Defense in Depth capacity) are needed for design feature
- Protection of Plant Workers (C) Functional Design Criteria (Personnel; Concept of Operations) 9
Technology-Inclusive Methodology Primary Safety Function What function(s)
(e.g., a barrier, cooling) are (limiting release of radioactive materials) needed to satisfy safety Primary (MHTGR example) criteria Safety Functions Additional Reactivity/Heat Chemical Heat Removal Generation Interactions What design features (e.g., a structure, system)
Design Features Design Features Design Features are provided to fulfill the safety function(s)
(and Human Actions)
Functional Functional What design criteria (e.g., leak rate, cooling Functional Design Criteria Design Criteria Design Criteria capacity) are needed for (Personnel; Concept of design feature Operations) 10
Modular High-Temperature Gas-Cooled Reactor (MHTGR) Example (Safety Functions)
Design Features Design Features Design Features Functional Design Functional Design Functional Design Criteria Criteria Criteria 11
Addressing Functions & Design Criteria (B) Safety Functions (C) Design Features (C) Functional Design Criteria 10 CFR 50, Appendix A General Design Criteria Quality Standards and Records 1 Design Bases for Protection Against Natural Phenomena 2 I. Overall Requirements: Fire Protection 3 Environmental and Dynamic Effects Design Bases 4 Sharing of Structures, Systems, and Components 5 Reactor Design 10 Reactor inherent Protection 11 Suppression of Reactor Power Oscillations 12 Instrumentation and Control 13 II. Protection by Multiple Fission Product Reactor Coolant Pressure Boundary 14 Reactor Coolant System Design 15 Barriers: Containment Design 16 Electric Power Systems 17 Inspection and Testing of Electric Power Systems 18 Control Room 19 Protection System Functions 20 Protection System Reliability and Testability 21 Protection System Independence 22 Protection System Failure Modes 23 III. Protection and Reactivity Control Separation of Protection and Control Systems Protection System Requirements for Reactivity Control 24 25 Systems: Malfunctions Reactivity Control System Redundancy and Capability 26 Combined Reactivity Control Systems Capability 27 Reactivity Limits 28 Protection Against Anticipated Operational Occurrences 29 12
Addressing Functions & Design Criteria (B) Safety Functions (C) Design Features Quality of Reactor Coolant Pressure Boundary 30 Fracture Prevention of Reactor Coolant Pressure Boundary 31 (C) Functional Design Criteria Inspection of Reactor Coolant Pressure Boundary 32 Reactor Coolant Makeup 33 Residual Heat Removal 34 Emergency Core Cooling 35 Inspection of Emergency Core Cooling System 36 Testing of Emergency Core Cooling System 37 IV. Fluid Systems: Containment Heat Removal 38 Inspection of Containment Heat Removal System 39 Testing of Containment Heat Removal System 40 Containment Atmosphere Cleanup 41 Inspection of Containment Atmosphere Cleanup Systems 42 Testing of Containment Atmosphere Cleanup Systems 43 Cooling Water 44 Inspection of Cooling Water System 45 Testing of Cooling Water System 46 Containment Design Basis 50 Fracture Prevention of Containment Pressure Boundary 51 Capability for Containment Leakage Rate Testing 52 Provisions for Containment Testing and Inspection 53 V. Reactor Containment: Systems Penetrating Containment 54 Reactor Coolant Pressure Boundary Penetrating Containment 55 Primary Containment Isolation 56 Closed Systems Isolation Valves 57 Control of Releases of Radioactive Materials to the 60 Environment Fuel Storage and Handling and Radioactivity Control 61 VI. Fuel and Radioactivity Control: Prevention of Criticality in Fuel Storage and Handling 62 Monitoring Fuel and Waste Storage 63 Monitoring Radioactivity Releases 64 13
Part 50 and Part 53 Comparing Licensing Frameworks
- Safety criteria o Same safety criteria in Parts 50 and 53 o Quantitative health objectives (QHOs) used in guidance under Part 50
- Design and Analyses o Design Basis Accidents (DBAs)
Part 50: Assessed using prescriptive, highly conservative analyses Including single failure criterion (SFC)
Part 53: Assessed methodically considering event frequencies and assuming only safety-related SSCs are available o Beyond Design Basis Events (BDBEs)
Part 50: Identified & assessed by largely ad-hoc, prescriptive approach with uncertainties addressed through conservatisms Part 53: Derived methodically using event frequencies with explicit consideration for uncertainties Including combinations of various equipment failures
- Special Treatment for Non-Safety-Related but Risk-Significant SSCs o Part 50: Ad-hoc (e.g., § 50.69 programs, Reliability Assurance Programs (RAP))
o Part 53: Systematic approach to control frequencies and consequences of the LBEs in relation to safety criteria 14
Second Iteration - Objectives
§ 53.200 Safety Objectives.
Each advanced nuclear plant must be designed, constructed, operated, and decommissioned to limit the possibility of an immediate threat to the public health and safety. In addition, each advanced nuclear plant must take such additional measures as may be appropriate when considering potential risks to public health and safety. These safety objectives shall be carried out by meeting the safety criteria identified in this subpart.
- Discussion o Generally aligns with requirements for content of technical specifications and regulatory treatment of non-safety systems o Addresses concerns related to tying tiers to authorities provided in the Atomic Energy Act (e.g., adequate protection and minimize danger to life or property) 15
Second Iteration - First Tier
§ 53.210 First Tier Safety Criteria.
(a) Public dose does not exceed Part 20 limit (0.1 rem) from normal plant operation (b) Provide design features and programmatic controls such that events with frequencies greater than once per 10,000 years meet the following (1) 2-hour dose below 25 rem at EAB (2) Duration dose below 25 rem at LPZ boundary
- Discussion o Maintains technical criteria from first iteration o Generally aligns with requirements for content of technical specifications and regulatory treatment of non-safety systems o Deleted paragraph (c) since the first tier criteria are no longer tied to adequate protection standard o Added existing footnote on 25 roentgen equivalent man (rem) as reference value o General note that staff assessing terminology (tiers) 16
Additional Discussion - First Tier
- Possible Applications of First Tier Safety Criteria o Minimally acceptable level of safety o Met by satisfying the safety functions needed for dose < 25 rem o Provides basis for safety classification of safety-related SSCs o Demonstration of meeting the first tier safety criteria supported by analyses of DBA o Provides basis for identifying SSCs needing protection against external events up to the design basis external hazard levels o Provides basis for identifying appropriate content of technical specifications (TS)
Reserved for the most significant safety requirements Necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety o May provide basis for staffing and operator licensing decisions o Greatest level of detail for information in licensing documents 17
Second Iteration - Second Tier
- Second Tier Safety Criteria FIRST ITERATION/SECOND ITERATION
§ 53.220 Second Tier Safety Criteria.
(a) Normal operations. Design features and programmatic controls must be provided for each advanced nuclear plant to ensure the estimated total effective dose equivalent to individual members of the public from effluents resulting from normal plant operation are as low as is reasonably achievable taking into account the state of technology, the economics of improvements in relation to the state of technology, operating experience, and the benefits to the public health and safety. Design features and programmatic controls must be established such that [to be reworded for consistency with 10 CFR part 20 and 40 CFR part 190].
(b) Unplanned events. Design features and programmatic controls must be provided to:
(1) Ensure plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address licensing basis events in accordance with
§ 53.240 and provide measures for defense-in-depth in accordance with § 53.250; and (2) Maintain overall cumulative plant risk from licensing basis events such that the risk to an average individual within the vicinity of the plant receiving a radiation dose with the potential for immediate health effects remains below five in 10 million years, and the risk to such an individual receiving a radiation dose with the potential to cause latent health effects remains below two in one million years.
18
Feedback - 2nd Tier, ALARA
- NRC Iteration: Maintained requirements for normal operations and occupational exposures to be ALARA Note that concerns related to ALARA and NRC reviews of design-related applications are also being addressed through the Advanced Reactor Content of Application Project with current drafts of Chapter 9 released to support stakeholder interactions:
... in lieu of providing detailed system descriptions and analysis of estimated effluent releases as required by 10 CFR 50.34, 50.34a, 52.47, and 52.79, an application may demonstrate compliance with the applicable regulations by describing a radiation protection program and an effluent release monitoring program that will ensure that effluent release limits will be met during normal operations for the life of the plant.
Information related to physical systems can be limited to general descriptions of layout and technologies used to limit the release of the various inventories of radioactive materials within the plant.
19
Feedback - 2nd Tier, QHOs
- QHOs o Proposal by some stakeholders to maintain QHOs as policy but exclude from rule Some concern over use of QHOs related to inclusion of requirement to perform PRA o Proposal by some stakeholders to use a metric other than QHOs as second tier Range of stakeholder views, from use of QHOs to use of cost-benefit assessment for second tier, which in NRC practice includes assessment against QHOs
- NRC Iteration: Maintained QHOs within the second tier safety criteria o The QHOs are a well-established measure used in NRC risk-informed decision making and are a logical performance metric to support the risk management approaches to operations that will be reflected in Subpart F, Operations.
o Note that using less defined criteria for the second tier would decrease the predictability of the regulations in terms of the desired graded approach (e.g., differentiation between SSCs that are safety related and non-safety related with special treatment) 20
Additional Discussion - Second Tier
- Possible Applications of Second Tier Safety Criteria o With first tier, ensures appropriate level of safety for long-term, risk-informed operations o Met by satisfying the safety functions for meeting QHOs o Demonstration of meeting the second tier safety criteria supported by systematic analyses (i.e., PRA) o Provides basis for identifying additional risk-informed requirements o Provides basis for identifying appropriate special treatment for non-safety related SSCs (e.g., functional design requirements & reliability) o Provides basis for enabling risk management approach to operations o May provide basis for staffing and operator licensing decisions o Enables appropriate level of detail in licensing basis documentation based on a risk-informed, function-oriented and performance-based approach 21
Second Iteration - Safety Functions
§ 53.230 Safety Functions (a) The primary safety function is limiting the release of radioactive materials from the facility and must be maintained during routine operation and for licensing basis events over the life of the plant.
(b) Additional safety functions supporting the retention of radioactive materials during routine operation and licensing basis eventssuch as controlling [reactivity], heat generation, heat removal, and chemical interactions--must be defined.
(c) The primary and additional safety functions are required to meet the first and second tier safety criteria and are fulfilled by the design features and programmatic controls specified throughout this part.
- Discussion (Safety Functions) o Maintains mention of fundamental safety functions as examples to maintain technology-inclusive framework (with potential use for multiple inventories of radionuclides within plants and possibly technologies such as fusion energy systems) o Reinforces general hierarchy of safety criteria, safety function, design feature, and functional design criteria.
22
Second Iteration - LBEs
§ 53.240 Licensing Basis Events Licensing basis events must be identified for each advanced nuclear plant and analyzed in accordance with § 53.450 to support assessments of the safety requirements in this subpart B. The licensing basis events must address combinations of malfunctions of plant SSCs, human errors, and the effects of external hazards ranging from anticipated operational occurrences to very unlikely event sequences with estimated frequencies well below the frequency of events expected to occur in the life of the advanced nuclear plant. The evaluation of licensing basis events must be used to confirm the adequacy of design features and programmatic controls needed to satisfy first and second tier safety criteria of this subpart and to establish related functional requirements for plant SSCs, personnel, and programs.
Licensing Basis Events - Light-Water Reactor (LWR) Summary ANSI/ANS-51.1-1983; nuclear safety criteria for the design of stationary pressurized water reactor plants (withdrawn 1989) 24
Licensing Modernization Project (LMP):
Event Selection & Analysis
- Introduction of an actual frequency-consequence curve as part of the regulatory process (vs. general relationship of decreased consequences expected for more frequent events) 25
Tabletop Exercise (MHTGR; Xe-100)
Report: ADAMS Accession No. ML18228A779 26
LMP: Event Selection & Analysis Anticipated Operational Occurrences (AOOs)
[Part 53 - AOOs]
Anticipated event sequences expected to occur one or more times during the life of a nuclear power plant, which may include one or more reactor modules. Event sequences with mean frequencies of 1x10-2/plant-year and greater are classified as AOOs. AOOs take into account the expected response of all SSCs within the plant, regardless of safety classification.
[Part 53 - Unlikely events]
Infrequent event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactor modules, but are less likely than AOOs. Event sequences with mean frequencies of 1x10-4/plant-year to 1x10-2/plant-year are classified as DBEs. DBEs take into account the expected response of all SSCs within the plant regardless of safety classification.
[Part 53 - Very unlikely events]
Rare event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactor modules, but are less likely than a DBE. Event sequences with mean frequencies of 5x10-7/plant-year to 1x10-4/plant-year are classified as BDBEs. BDBEs take into account the expected response of all SSCs within the plant regardless of safety classification.
27
LMP: Required Safety Functions Required Safety Function (RSF): A PRA Safety Function that is required to be fulfilled to maintain the consequence of one or more DBEs or the frequency of one or more high-consequence BDBEs inside the F-C Target Provides connection to Safety-Related Classification Note - in Part 53, RSFs would translate to those functions needed to address first tier safety criteria 28
RSF Example
- MHTGR RSFs Required Safety Functions 29
Design Basis Accidents DBAs
[Part 53 - DBAs]
Postulated event sequences that are used to set design criteria and performance objectives for the design of Safety Related SSCs. DBAs are derived from DBEs based on the capabilities and reliabilities of Safety-Related SSCs needed to mitigate and prevent event sequences, respectively. DBAs are derived from the DBEs by prescriptively assuming that only Safety Related SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose limits.
30
Second Iteration - DiD
§ 53.250 Defense in Depth Measures must be taken for each advanced nuclear plant to ensure appropriate defense in depth is provided to compensate for uncertainties such that there is high confidence that the safety criteria in this subpart are met over the life of the plant. The uncertainties to be considered include those related to the state of knowledge and modeling capabilities, the ability of barriers to limit the release of radioactive materials from the facility during routine operation and for licensing basis events, and those related to the reliability and performance of plant SSCs, personnel, and programmatic controls. No single engineered design feature, human action, or programmatic control, no matter how robust, should be exclusively relied upon to meet the safety criteria of § 53.220(b) or the safety functions defined in accordance with § 53.230.
- Discussion (DiD) o Maintains defense in depth within Subpart B because of historical and continued importance of its role in addressing risk o Parts 50/52 do not include a similar section because the defense-in-depth philosophy is incorporated into prescriptive technical requirements for light-water reactors o Possibility that this section could be addressed within Subpart C can be considered as part of the later review of the technical requirements o Reflects possible crediting of inherent characteristics within the design and analysis for advanced reactors and the reduced uncertainties associated with such characteristics 31
Second Iteration - Protection of Plant Workers
§ 53.260 Protection of Plant Workers (a) Design features and programmatic controls must exist for each advanced nuclear plant to ensure that radiological dose to plant workers does not exceed the occupational dose limits provided in subpart C to 10 CFR part 20.
(b) As required by Subpart B to 10 CFR part 20, design features and programmatic controls must, to the extent practical, be based upon sound radiation protection principles to achieve occupational doses that are as low as is reasonably achievable.
- Discussion (Protection of Plant Workers) o Maintains the protection of plant workers within Subpart B to capture occupational exposures within the high-level safety requirements o Changed to refer to part 20, as suggested by stakeholders Note that ALARA is not only a long-standing requirement by Atomic Energy Commission/NRC (including maintaining in Part 20 rulemaking) but also is addressed in U.S. Environmental Protection Agency Federal Guidance for Radiation Protection 32
Subpart C Design and Analysis Preliminary Language 33
Subpart C - Design and Analysis
- Design Features
- Functional Design Criteria for First Tier Safety Criteria
- Comparable to Principal Design Criteria for Safety-Related SSCs
- Functional Design Criteria for Second Tier Safety Criteria
- Provides Design Criteria for Safety Significant Non-Safety-Related SSCs
- Functional Design Criteria for Protection of Plant Workers
- Design Requirements
- Analysis Requirements
- Role of PRA
- Safety Categorization and Special Treatment
- Application of Analytical Safety Margins to Operational Flexibilities
- Design Control Quality Assurance
- Design and Analyses Interfaces 34
Design-Related Discussions SFC vs Reliability Criterion Part 53 PRA Required; Reliability Assurance through TS/RAP Subpart F RG 1.233 The staff finds that the NEI 18-04 methodology, including (Licensing assessments of event sequences and DiD, obviates the Modernization) need to use the single-failure criterion (SFC) as it is applied to the deterministic evaluations of AOOs and DBAs for (SECY-19-0117)
LWRs.
SRM-SECY-19-0036 The staff should apply risk-informed principles when strict, (Application of the Single prescriptive application of deterministic criteria such as the Failure Criterion to SFC is unnecessary to provide for reasonable assurance of NuScale IAB Valves) adequate protection of public health and safety.
SECY-03-0047 The SFC would be replaced with a reliability criterion and (Policy Issues Related to the event scenarios identified in the PRA would be Licensing Non-Light- examined against this criterion.
Water Reactor (NLWR)
Designs)
Note that Issue 4 in SECY-03-0047 also described SRM dated 6/26/2003 probabilistic event selection and safety classification 35
The SFC
- The SFC has the direct objective of promoting reliability through the enforced provision of redundancy in those systems which must perform a safety-related function
- In SECY 77-439 (ML060260236), the staff critiqued the SFC at the request of the Commission:
- The SFC has served well in its use as a licensing review tool to assure reliable systems as one element of the defense in depth approach to reactor safety.
- The SFC is just one of several tools applied in systems design and analysis to promote reliability of the systems which are needed in a nuclear power plant for safe shutdown and cooling, and for mitigation of the consequences of postulated accidents.
It is not sufficient by itself.
- The SFC was developed without the benefit of numerical assessments on the probabilities of component or system failure.
- The Reactor Safety Study (WASH-1400, the first nuclear plant PRA) also pointed out that factors such as systems interactions, multiple human errors, and maintenance and testing requirements also have an influence on reliability. Such factors fall outside the scope of the SFC, and supplementary methods must be utilized In their study.
- It is expected that probabilistic methods of the type used in the Reactor Safety Study will gradually come into increasing use and supplement the SFC.
- See also the discussions in SECY-03-0047, SECY-05-0138, SECY-19-0036, SECY-19-0117 and related SRMs 36
Codes and Standards
§ 53.440 Design Requirements.
(a) The design features required to meet the first and second tier safety criteria defined in
§§ 53.210 and 53.220 shall be designed using generally accepted consensus codes and standards wherever applicable.
Preliminary Definition (Subpart A): Consensus code or standard means any technical standard (1) developed or adopted by a voluntary consensus standard body under procedures that assure that persons having interests within the scope of the standard that are affected by the provisions of the standard have reached substantial agreement on its adoption, (2) formulated in a manner that afforded an opportunity for diverse views to be considered, and (3) designated by the standards body as such a standard for the safe design, manufacture, construction, or operation of nuclear power plants.
- Discussion (Codes and Standards)
- Preliminary language encourages use of consensus codes and standards as required by the National Technology Transfer and Advancement Act.
- Recognizes variety of technologies and designs as well as stated desire of some stakeholders to adopt standards outside of typical LWR standards development organizations (e.g., ISO or other international standards).
- Considering using NRC endorsement of guidance documents versus incorporation of standards into the regulations.
- Capture of acceptable standards in guidance increases efficiency by avoiding routine rulemakings related to the revision of incorporated standards in the regulations.
37
Second Iteration - Analysis (PRA)
§ 53.450 Analysis Requirements (a) Requirement to have a probabilistic risk assessment. A probabilistic risk assessment (PRA) of each advanced nuclear plant [reminder - plant definition to include multi-module and multi-source] must be performed to identify potential failures, degradation mechanisms, susceptibility to internal and external hazards, and other contributing factors to unplanned events that might challenge the safety functions identified in § 53.230 and to support demonstrating that each advanced nuclear plant meets the second tier safety criteria of § 53.220(b).
- Discussion (PRA) o Maintains requirement in Part 53 for PRA consistent with evolution of risk-informed approaches but provide alternatives to PRA for design and analysis processes (paragraph (b)) and to support the licensing and regulatory programs being developed in subsequent subparts o Staff is engaged in ongoing discussions on how to ensure the level of effort required for a PRA is commensurate with the complexity of the subject reactor design while also ensuring possible deployment of advanced reactors poses no undue risk to public health and safety.
38
Past and Present Uses of the PRA
- Identify severe accident vulnerabilities and to provide insights which support the conclusion that the plant design, construction, and operation provides reasonable assurance no undue risk to public health and safety.
- Demonstrate that the plant meets the Commissions safety goals.
- Support the environmental review required by 10 CFR Part 51, specifically, the evaluation of severe accident mitigation design alternatives:
- RG 4.2, Preparation of Environmental Reports for Nuclear Power Stations, Rev. 3, September 2018
- COL-ISG-029, Environmental Considerations Associated with Micro-reactors, October 28, 2020
- For applications based on the LMP guidance, the PRA is used to select licensing basis events, classify SSCs, and to inform the DiD evaluation.
39
Past and Present Uses of the PRA (Cont'd)
- For applications not based on the LMP guidance, the PRA may be used to support the process used to demonstrate whether the regulatory treatment of non-safety systems (RTNSS) is sufficient and, if appropriate, identify the SSCs included in RTNSS.
- The results and insights of the PRA are used to identify and support the development of specifications and performance objectives for the plant design, construction, inspection, and operation, such as:
- Inspection, testing, analysis, acceptance criteria,
- TS, and
- Combined operating license action items and interface requirements.
- The PRA may be used to support various voluntary risk-informed applications (e.g., risk-informed inservice inspection) that may be included in the licensing application.
- The PRA may be used to inform the scope of staffs review; see SRM-COMGBJ-10-0004/COMGEA-10-0001 (ML102510405).
- The results and insights of the PRA are used to support the reactor oversight program.
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Searching for Initiating Events (Adapted from the NLWR PRA Standard)
- Identify initiating events that:
- Challenge normal plant operation (when plant is at-power) or the ability to sustain safe shutdown or low-power conditions (when not at-power), and
- Require successful mitigation to prevent a release of radioactive material.
- Use a structured, systematic process that accounts for plant-or design-specific features, such as:
- Master logic diagrams
- Heat balance fault trees
- Process hazards analysis
- Failure modes and effects analysis
- Analyze operating procedures and practices.
- Review existing lists of known initiators applicable to the specific reactor type and design.
41
Searching for Initiating Events (Cont'd)
(Adapted from the NLWR PRA Standard)
- Consider external hazards (e.g., seismic), including initiating events caused by a combination of hazards (e.g., seismically induced fires).
- Review operating experience, including similar plants.
- Perform a systematic evaluation of each system down to the subsystem or train level and including support systems in each modeled plant operating state.
- Include initiating events resulting from multiple failures if the equipment failures result from a common cause.
- Interview resources knowledgeable in plant design or operation.
- Include initiators that impact two or more sources of radioactive material 42
Addressing Lack of Operating Experience Type of Data/Information Methods Internal initiating event frequencies
- Many can be estimating using LWR or relevant non-nuclear information Component failure rates
- Bayesian estimation methods
- Formal expert elicitation Common-cause failures (CCFs)
- Use existing CCF models (e.g., alpha factors)
- Use existing generic information derived from LWR experience Test/maintenance availabilities
- Use component failure rates
- Controlled by technical specifications (surveillance test intervals and allowed outage times)
Human error probabilities
- Does not require design-specific operating experience External hazard frequencies
- Use existing methods External hazard fragilities 43
Addressing Lack of Operating Experience (Cont'd)
- PRA provides a framework for assessing uncertainties:
- Parametric uncertainties
- Modeling uncertainties
- Completeness uncertainties
- PRA helps to put uncertainties into perspective.
- Which events contribute to the overall uncertainty?
- Are these events also risk significant?
44
Second Iteration - Analysis (Use of PRA)
§ 53.450 Analysis Requirements (b) Requirement to use PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof to:
- Determine LBEs
- Support safety classification of SSCs
- Evaluate defense in depth
- Discussion (Use of PRA) o Change intended to support alternative approaches to a PRA o Worded in terms of generally accepted to support possible standards or other guidance documents o The use of guidance, Part 53 rule language, or revisions to Part 50 are being explored as possible ways to accommodate deterministic approaches for performing design and analysis 45
Second Iteration - Analysis Requirements (c - g)
§ 53.450 Analysis Requirements (c) Maintenance and upgrade of analyses (d) Qualification of analytical codes (e) Analyses of LBEs (added)
(f) Analysis of DBAs (g) Other required analyses
- Discussion (Analysis Requirements) o Clarification of maintenance and upgrading of analyses (referring to codes and standards) o Maintain placeholder for other required analyses to address fire protection, aircraft impact, and specific beyond design basis accidents.
46
Second Iteration - Analysis Requirements (c - g)
§ 53.450(e) Analyses of licensing basis events [New sub-paragraph]
(e) Analyses of licensing basis events. Analyses must be performed for licensing basis events ranging from anticipated operational occurrences to very unlikely event sequences with estimated frequencies well below the frequency of events expected to occur in the life of the advanced nuclear plant. The licensing basis events must be identified using insights from a PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof to systematically identify and analyze equipment failures and human errors. The analyses must address event sequences from initiation to a defined end state and demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the second tier safety criteria of § 53.220(b) and provide defense in depth as required by § 53.250.
- Discussion (Analyses of LBEs) o Section added to clarify requirements for LBEs, including analysis from initiation to a defined end state o Staff considering further clarification for anticipated operational occurrences in terms of acceptance criteria beyond QHOs and defense in depth 47
Second Iteration - Analysis Requirements (c - g)
§ 53.450 (f) Analysis of design basis accidents (f) Analysis of design basis accidents. The analysis of licensing basis events required by § 53.240 and § 53.450(e) must include analysis of a set of design basis accidents that address possible challenges to the safety functions identified in accordance with § 53.230. Design basis accidents must be selected from those unanticipated event sequences with an upper bound frequency of less than one in 10,000 years as identified using insights from a PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof to systematically identify and analyze equipment failures and human errors. The events selected as design basis accidents should be those that, if not terminated, have the potential for exceeding the safety criteria in § 53.210(b). The design-basis accidents selected must be analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the safety-related SSCs identified in § 53.460 and human actions addressed by § 53.8xx (reference to concept of operations sections of Subpart F) are available to perform the safety functions identified in accordance with § 53.230. The analysis must conservatively demonstrate compliance with the safety criteria in § 53.210(b).
- Discussion (DBAs) o Revised to clarify that analysis is to address sequences from initiation to a safe stable end state.
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Second Iteration - Safety Classification
§ 53.460 Safety Categorization and Special Treatment (a) SSCs and human actions must be classified according to their safety significance.
The categories must include Safety Related (SR), Non-Safety Related but Safety Significant (NSRSS), and Non-Safety Significant (NSS), as defined in subpart A of this part.
- Discussion o Editorial changes to remove material duplicating preliminary rule language in other sections o Maintaining for now the specific categories of safety related, non-safety related but safety significant, and non-safety significant 49
Second Iteration - Analytical Margins and Operating Flexibilities
§ 53.470 Application of Safety Margins to Operational Flexibilities (No Change) Where an applicant or licensee so chooses, design criteria more restrictive than those defined in § 53.220(b) may be adopted to support operational flexibilities (e.g., emergency planning requirements under Subpart F of this part). In such cases, applicants and licensees must ensure that the functional design criteria of § 53.420(b), the analysis requirements of § 53.450, and identification of special treatment of SSCs and human actions under
§ 53.460 reflect and support the use of alternative design criteria to obtain additional analytical safety margins. Licensees must ensure that measures taken to provide the analytical margins supporting operational flexibilities are incorporated into design features and programmatic controls and are maintained within programs required in other Subparts.
- Discussion o No change; Released related requirements in Subpart F to th support public meeting on May 6 50
Feedback - Design Control Quality Assurance and Design Interfaces First Iteration
§ 53.480 Design Control Quality Assurance
§ 53.490 Design Interfaces
- Questions/comments on quality assurance and design interfaces o Many stakeholders reserving comments pending release of other subparts
- Discussion o No change; Released related requirements in Subpart F to support public meeting on May 6th 51
Feedback - Non-Radiological Hazards
- Non-Radiological Hazards o Some ACRS members noted inclusion of non-radiological hazards should be considered in Part 53, such as chemical releases.
Staff has this issue under consideration and recognizes existing frameworks for addressing this multi-jurisdictional topic Does ACRS have feedback on this topic that could inform the Staffs ongoing considerations?
52
Final Discussion and Questions 53
Acronyms and Abbreviations ACRS Advisory Committee on Reactor Safeguards DiD Defense in depth ADAMS Agencywide Document Access EAB Exclusion area boundary Management System EP Emergency planning AEA Atomic Energy Act EPA U.S. Environmental Protection Agency ALARA As low as reasonably achievable F-C Frequency consequence ANS American Nuclear Society FMEA Failure modes and effects analysis AOO Anticipated operational occurrence FW Steam generator feedwater pump trip ASME American Society of Mechanical Engineers BDBEs Beyond design basis events HPB Helium pressure boundary CCF Common cause failure IAB Intake air bypass CFR Code of Federal Regulations ISO International Standards Organization ITAAC Inspection, test, analyses, acceptance CR Control rod withdrawal criteria CT Circulator trip LBEs Licensing basis events DBAs Design basis accidents LD Large helium depressurization DG Draft guidance LF Loss of primary flow 54
Acronyms and Abbreviations LMP Licensing modernization project PC Plant condition LO Loss of offsite power PPC Porcelain polycarbonate PRA Probabilistic risk assessment LPZ Low-population zone LWR Light-water reactor QHO Quantitative health objective MD Medium helium depressurization RAP Reliability assurance program MHTGR Modular high-temperature gas-cooled Rem Roentgen equivalent man reactor ROP Reactor oversight program NEI Nuclear Energy Institute NEIMA Nuclear Energy Innovation and RSF Required safety function Modernization Act RT Reactor trip NLWR Non-light-water reactor RTNSS Regulatory treatment of non-safety systems NRC U.S. Nuclear Regulatory Commission SAR Safety analysis report NSRSS Non-safety related but safety significant SD Small helium depressurization NSS Non-safety significant SDO Standard development organization PAG Protective action guide SFC Single-failure criterion 55
Acronyms and Abbreviations SG Steam generator rupture SR Safety related SSCs Structures, systems, components TS Technical specifications TT Turbine trip 56
BACKUP SLIDES 57
Part 53 Rulemaking Schedule Milestone Schedule Major Rulemaking Activities/Milestones Schedule Public Outreach, ACRS Interactions and Present to April 2022 Generation of Proposed Rule Package (11 months)
Submit Draft Proposed Rule Package to May 2022 Commission Publish Proposed Rule and Draft Key Guidance October 2022 Public Comment Period - 60 days November and December 2022 Public Outreach and Generation of Final Rule January 2023 to February 2024 Package (14 months)
Submit Draft Final Rule Package to Commission March 2024 Office of Management and Budget and Office of July 2024 to September 2024 the Federal Register Processing Publish Final Rule and Key Guidance October 2024 58
Integrated Approach Siting near densely populated Functional areas EP for SMRs Containment and ONTs Licensing Modernization Project Insurance and Liability Environmental Reviews Consequence Based Security 59
Presenting PRA Results 60
Cumulative Risk Metrics
- QHOs in the Commissions Safety Goal Policy Statement
- The risk to an average individual in the vicinity of a nuclear power plant [1 mile] of prompt fatalities that might results from reactor accidents should not exceed 0.1% of the sum of prompt fatality risk resulting from other accidents to which members of the U.S. population are generally exposed [5E-7/y].
- The risk to the population in the area near a nuclear power plant
[10 miles] that might result from nuclear power plant operation should not exceed 0.1% of the sum of cancer fatality risks resulting from all other causes [2E-6/y].
- Compare mean risks to QHOs, and consider the uncertainties
- Basis: NUREG-0880, Safety Goals for Nuclear Power Plant Operation, Rev. 1, ML071770230, May 1983.
- LMP: The total mean frequency of exceeding a site boundary dose of 100 mrem < 1/plant-year (based on 10 CFR 20).
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Large Release Frequency (LRF)
- In its safety goal policy statement, the Commission proposed a general performance guideline for further staff examination:
- The overall mean frequency of a large release of radioactive materials to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation
- Rationale as explained by Forrest Remick (former Director of Office of Policy Evaluation, former ACRS member, and former Commissioner) in a memorandum dated 3/2/1993 (ML051660709) to James Taylor (former EDO):
- The ACRS wanted to include a goal for conditional containment failure probability (CCFP) < 0.1
- The LRF goal was developed to break the deadlock between the staff and ACRS
- In SRM-SECY-89-102 (ML051660712), the Commission made clear that LRF applies to all reactor designs (LWRs and NLWRs).
- As discussed in SECY-93-138, the staff abandoned efforts to anchor LRF to the QHOs (LRF is more conservative).
- There is no NRC definition for LRF; Part 52 applicants have been allowed to propose various definitions.
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Large Release vs.
Large Early Release
- JCNRM definition of large release (approved 4/2/2021): The release of airborne fission products to the environment such that there are significant off-site impacts. Large release and significant off-site impacts may be defined in terms of quantities of fission products released to the environment, status of fission product barriers and scrubbing, or dose levels at specific distances from the release, depending on the specific analysis objectives and regulatory requirements.
- RG 1.200 implied definition of large early release: A rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is the potential for early health effects. (Such accidents generally include unscrubbed releases associated with early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation.)
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Core-Damage Frequency (CDF) and Large Early Release Frequency (LERF)
- For large LWRs:
- Used when developing RG 1.174 (late 1990s)
- Technical basis documented in NUREG-1860, Appendix D (based on NUREG-1150 results)
- In SRM-SECY-12-0081,the Commission approved the staff's recommendation that new reactors transition from LRF to LERF at or before initial fuel load.
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CCDF Representation of Risk
- Used in traditional PRAs (e.g., WASH-1400, NUREG-1150)
- Considered during development of NUREG-1860, Feasibility Study for a Risk-Informed and Performance-Based Regulatory Structure for Future Plant Licensing
- Deferred - how to establish the acceptance criterion?
- Discussed in ACRS letter dated September 26, 2007
- Public comment on DG-1353 [RG 1.233] by former ACRS Member Rich Denning and Vinod Mubayi (one of the authors of NUREG-1860) recommended the development of a CCDF criterion in lieu of the frequency-consequence target:
- Comment: ML19158A457
- Staff response: ML20091L696
- Discussed at ACRS Future Plant Design Subcommittee meeting held July 20, 2020 65
Frequency-Consequence Plot
- Uses include:
- MHTGR pre-application (1989)
- NUREG-1860 (2007)
- NGNP Licensing Strategy (2008)
- NEI 18-04 (2019)
- In NEI 18-04:
- The F-C Target is used as a tool to identify risk-significant event sequence families and SSCs
- The F-C Target is not an acceptance criterion!
66